Text

Advises of No Known Public Correspondence or Irreversible Impact Associated W/Issuance of Amend 12 to License DPR-77
	ML20054D406
Person / Time
Site:	Sequoyah
Issue date:	03/24/1982
From:	Stahle C; Office of Nuclear Reactor Regulation
To:	Tedesco R; Office of Nuclear Reactor Regulation
Shared Package
ML20054D398	List: ML20054D397; ML20054D400; ML20054D401; ML20054D404; ML20054D406;
References
	NUDOCS 8204220674
	Download: ML20054D406 (1)
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{{#Wiki_filter:T5 75 h Tl, T 4 A W w MARCH 2 4 1982 Docket No. 50-327 MEMORANDUM FOR: Robert L. Tedesco, Assistant Director for Licensing Division of Licensing THRU: Elinor G. Adensam, Chief Licensing Branch No. 4 Division of Licensing FROM: Carl Stahle, Project Manager Licensing Branch Ho. 4 Division of Licensing

SUBJECT:

ISSUANCE OF AMENDMENT NO.12 TO FACILITY OPERATING LICENSE DPR-77 SEOUOYAH HUCLEAR PLANT, UNIT 1 There is no known public correspondence or irreversible impact associated with the issuance of the subject anendment. " Original Signed By: WN" f Carl Stahle, Project Manager ( f. M Licensing Branch No. 4 Division of Licensing cu:"' l.A;DL ;jB.#4.......[L.*g..#.4. ..... 0 M. 8204220674 820325 I,Adens am.. PDR ADOCK 05000327 P, PD,R,,__,,,,,,2/3).//82.... w:c ronu si8 io.eoinacu o24o - OFFICIAL RECORD COPY

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A'4ENDMENT NO.12 TO FACILITY OPERATING LICENSE DPR SE0J0YAH N'JCLEAR PLANT, UNIT NO.1 DISTRIBUTION w/ enclosures: bec w/ enclosures: cket No. 50-327 NRC PDR / A LB f4 r/f Local PDR s g$pp # C. Stable NSIC 2j M. Duncan TERA f.O 5 I&E (5) A. Rosenthal, ASLAE 2 Yg E. Ketchen, OELD ASLBP W gh D 6 G. Deegan (4) ACRS (16) v E. Adensam DMB - 10 c3, s MPA A / R. Diggs, DE D c) D. Eisenhut R. Purple R. Tedesco T. Novak F. Miralgia B. J. Youngblood J. Miller A. Schwencer S. Hanauer R. Vollmer R. Mattson H. Thompson NMSS M. Virgilio

Attachment to License Amendment No. 12. Facility Operating License No. DPR-77 Docket No. 50-327 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified hy Amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness. Amended Overleaf Page Page 1-1 4 1-2 1-3 1-4 1-5 1-6 1-7 1-8 2-2 2-1 2-4 2-3 2-7 3/4 1-3 3/4 1-4 3/4 1-8 3/4 1-7 3/4 2-2 3/4 2-1 3/4 2-7 3/4 2-8 3/4 2-9 3/4 2-10 3/4 3-1 3/4 3-2 3/4 3-4 3/4 3-5 3/4 3-6 3/4 3-7 3/4 3-10 3/4 3-9 ~ 3/4 3-11 3/4 3-12 3/4 3-13 3/4 3-14 3/4 3-18 3/4 3-17. 3/4.3-20 3/4 3-19 3/4 3-23 3/4 3-24 3/4 3-25 3/4 3-26 3/4 3-27 4 3/4 3-28 3/4 3-31 3/4 3-32 3/4.3-35 3/4 3-36 3/4'3-37 3/4 3-38 3/4 3-40 3/4 3-39 1 w v-v n =,, - - - w Amended Overleaf Page Page 3/4 3-41 3/4 3-42 3/4 3-43 3/4 3-44 3/4 3-45 3/4 3-46 3/4 3-55 3/4 3-56 3/4 3-57 3/4 3-58 3/4 3-60 3/4 3-59 3/4 3-65 3/4 3-66 3/4 3-73 3/4 3-74 3/4 3-80 3/4 3-79 3/4 4-la 3/4 4-1 3/4 4-2 3/4 4-2a 3/4 4-2b 3/4 4-4a 3/4 4b 3/4 4-5 3/4 4-6 3/4 4-13 3/4 4-14 3/4 4-15 3/4 4-15a 3/4 4-23 3/4 4-24 3/4 4-23a 3/4 5-1 3/4 5-2 3/4 5-13 3/4 6-1 3/4 6-2 3/4 6-7 3/4 6-8 3/4 6-16 3/4 6-15 3/4 6-17 3/4 6-18 3/4 6-24 3/4 6-23 3/4 6-33 3/4 6-34 3/4 7-5 3/4 7-6 3/4 7-11 3/4 7-12 3/4 7-13 3/4 7-14 3/4 7-15 3/4 7-16 3/4 7-17 3/4 7-18 3/4 7-19 3/4 7-20 3/4 7-21 Amended Overleaf Page Page 3/4 7-22 '3/4 7-23 3/4 7-24 3/4 7-25 3/4 7-26 3/4 7-27 3/4 7-28 3/4 7 3/4 7-30 3/4 7-31 3/4 7-32 3/4 7-33 3/4 7-34 3/4 7-35 3/4 7-36 3/4 7-37 3/4 7-38 3/4 7-39 3/4 7-40 3/4 7-41 3/4 8-3 3/4 8-4 3/4 8-5 3/4 8-6 3/4 8-7a 3/4 8-9 3/4 8-10 3/4 8-12 3/4 8-11 3/4 8-13 3/4 8-14 3/4 8-13a 3/4 8-35 3/4 8-36 3/4 8-37 3/4 8-38 3/4 9-1 3/4 9-2 3/4 9-4 3/4 9-3 3/4 9-6 3/4 9-5 3/4 9-8a 3/4 10-1 3/4 10-2 3/4 11-9 3/4 11-10 3/4 11-11 3/4 11-12 3/4 11-15 3/4 11-16 B3/4 2-5 B3/4 2-6 B3/4 4-1 B3/4 4-2 83/4 7-3 B3/4 7-5 Amended Overleaf Page Page B3/4 7-6 B3/4 8-1 B3/4 8-2 B3/4 11-5 5-5 5-6 6-6 6-5 6-7 6-15b 6-16 6-22 6-21 6-23 6-24 In addition, some page numbers have been changed due to the deletion of the following pages: 3/4 6-25b 3/4 7-14 3/4 7-31 3/4 7-32 3/4 7-33 3/4 7-34 3/4 7-35 3/4 7-36 3/4 7-36a

1.0 DEFINITyrys DEFINED TERMS 6 The defined terms of this' section appear in capitalized type 2nd a : O!' mble throughout these Technical Specifications. ACTION 1.1 ACfl0N shall be that part of a Specification which prescribes rec t 'i measures required under designated conditions. AXIAL Fl.UX DIFFERENCE l 1.2 AKj AL llVX Dill Ef:fhCE shall be the dif ference in normalized flux sig,1ais I betwean the tcp and bottom halves of a two section excore neutron detector. t CHANNE L.CALIi! RATIO!j 1.3 A CuffuFl. cal iUJAT10N shall be t he adjystpent, as nccessary, of the-channel nutput. such thet it' responds uiti. the.necessary range and v.ruracy to known values of the pararreter which the channel taonitors. The CH'MFL CALIERA-T ION slu11 encucpass. thy entire channel incl,uding. the' sensor and alana and/or [ trip fuor tions, and :, hall incleb th: CH";Nf1 1;UNCT!GWL lthf. ihe CNiNFL i CALIDPla ION may he perfeired by any series of sec,uential, overlapping er total I chann"1 steps such that the untire channel is calibratod. CHANNEL.CHFCK.. -

1. 4 A Cia.NNEL C!!ECK ah ill be the qualitative assessnnt of ch mnel behavior i

during operation by rkervation. fhis determination shall include, where possible, co.rprison of the c' ;nnel indication rad /oc status sith, ther indica-tions and/or st atus derived frem in4erdent instrument cf mnels cmuring the same n m vter. CIIANNEL FUNCriONM ltST l 1.5 A CHANNEL IUNCfiONAL lESI shall Le: Analog cikanels - tie injection of a simulated signal into the a. channel as close to the sensor as practicable to trtify CYERG.II.ITY including alzab and/,cr t' rip functions. b. Bistable channels - the injection of a simulated signal into t he I

ensor to verify G?[PAB!l.liY including alana and/or trip innctions.

t SEQUOYAll - UNIT 1 1-3 A=cr hent No. 12

DEFINITIONS CONTAINMENT INTEGRITY 1.6 CONTAINMENT INTEGRITY shall exist when: a. All penetrations required to be closed during accident conditions are either: 1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2) Clcsed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-2 of Specification 3.6.3. i b. All equipment hatches are closed and sealed. Each air lock is OPERABLE pursuant to Specification 3.6.1.3, c. d. The containment leakage rates are within the limits of Specification 3.6.1.2, and The sealing mechansim associated with each penetration (e.g., e. welds, bellows, or 0 rings) is OPERABLE. CONTROLLED LEAKAGE 1.7 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals. CORE ALTERATION 1.8 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with'the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position. DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." SEQUOYAH - UNIT 1 1-2 Amendment No. 12

1 l DEFINITIONS F - AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. ENGINEERED SAFETY FEATURE RESPONSE TIME 1.11 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF' actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. FREQUENCY NOTATION 1.12 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2. GASE0US RADWASTE TREATMENT SYSTEM 1.13 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to redace radioactive gaseous ef fluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be: Leakage (except CONTROLLED LEAKAGE) into closed systems, such a. as pump seal or valve packing leaks that are captured and. conducted ~ to a sump or collecting tank, or b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or Reactor coolant system leakage through a steam generator to c. the secondary system. j 1 'SEQUOYAH - UNIT'1 1-3 Amendment No. 12

DEFINITIONS OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.15 The DFFSITE DOSE CALCULATION MANUAL shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous-and liquid effluents and in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints. It shall also contain the radiological environmental monitoring program. OPERABLE - OPERABILITY 1.16 A system, subsystem, train, or component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power souice, cooling or seal water, lubrication or other auxiliary -equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s). OPERATIONAL MODE - MODE 1.17 An OPERATIONAL NODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactcr coolant temperature specified in Table 1.1. PHYSICS TESTS 1.18 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission. PRESSURE BOUNDARY LEAKAGE 1.19 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. PROCESS CONTROL PROGRAM (PCP) 1.20 The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION of radioactive wastes from liquid systems is assured. PURGE - PURGING 1.21 PURGE or PUkGING'is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. i s SEQUOYAH - UNIT 1 1-4 Amendment No. 12

DEFINITIONS QUADRANT POWER TILT RATIO 1.22 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs,si.ichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average. RATED THERMAL POWER 1.23 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor' coolant of 3411 MWt. REACTOR TRIP SYSTEM RESPONSE TIME-1.24 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage. REPORTABLE OCCURRENCE 1.25 A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specifications 6.9.1.12 and 6.9.1.13. SHIELD BUILDING INTEGRITY 1.26 SHIELD BUILDING INTEGRITY shall exist when: The door in each access opening is closed except when the a. access opening is being used for normal transit entry and exit. b. The emergency gas treatment system is OPERABLE. The sealing mechanism associated with each penetration (e.g., c. welds, bellows or 0 rings) is OPERABLE. SHUTDOWN MARGIN 1.27 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subtritical or would be subcritical from its present condition j assuming all full' length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. 1 1 SEQUOYAH - UNIT 1 1-5 Amendment No. 12 i

l I DEFINITIONS i SOLIDIFICATION 1.28 SOLIDIFICATION shall be the conversion of radioactive wastes from liquid l systems to a uniformly distributed, monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides i (free-standing). l SOURCE CHECK 1.29 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source. i STAGGERED TEST BASIS I l 1.30 A STAGGERED TEST BASIS shall consist of: l A test schedule for n systems, subsystems, trains or other designated a. components obtained by dividing the specified test interval into n equal subintervals, i l b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval. THERMAL POWER 1.31 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. UNIDENTIFIED LEAKAGE l 1.32 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. l VENTILATION EXHAUST TREATMENT SYSTEM l 1.33 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. f VENTING 1.34 VENTING is the ct 4 olled process of discharging air or gas from a confinement to maintai' umperature, pressure, humidity, concentration or r other operating condition, in such a manner that replacement air or gas is nct'provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. SEQUOYAH - UNIT 1 1-6 Amendment No. 12

DEFINITIONS TABLE 1.1 OPERATIONAL MODES REACTIVITY % RATED AVERAGE COOLANT MODE CONDITION, K THERMAL POWER

TEMPERATURE eff 1.

POWER OPERATION > 0.09 > 5% > 350 F ~ 2. STARTUP > 0.99 5 5% > 350 F 3. HOT STANDBY < 0.99 0 > 350 F 4. HOT SHUTOOWN < 0.99 0 350 F > T "9 > 200*F 5. COLD SHUTDOWN < 0.99 0 < 200 F 6. REFUELING ** 5 0.95 0 < 140 F Excluding decay heat.

- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

l ? SEQUOYAH - UNIT 1 1-7 Amendment No. 12 t -e'-Q w -Q-.we s ,M e eh MEm

DEFINITIONS TABLE 1.2 FREQUENCY NOTATION l NOTATION FREQUENCY S At least once per 12 hours. O At least once per 24 hours. W At least once per 7 days. M At least once per 31 days. Q At least once per 92 days. SA At least once per 184 days. R At least once per 18 months. S/U Prior to each reactor startup. P Completed prior to each release. N.A. Not applicable. SEQUOYAH - UNIT 1 1-8

2.0 SAFETY" LIM 1TS AND LIMITING SAFETY SYSTEM SETTINGS ,. v 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the hig' hest operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figures 2.1-1 and 2.1-2 for 4 and 3 loop operation, respectively. i 6 ~ APPLICABILITY: MODES 1 and 2. ACTION: i Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour. REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig. APPLICABILITY: MODES 1, 2, 3, 4 and 5. ACTION: 1 MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with 'the Reactor Coolant System pressure within its limit within I hour. MCDES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes. SEQUOYAH - UNIT 1 2-1 m se ,e. g =e.-- e e. p .ee w

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Figure 2.1 1. Reactor Core Safety Limit-Four Loops in Operation SEQUOYAH - UNIT 1 2-2 Amendment No. 12

(. Figure 2.1-2 Reactor Core Safety Limit - Three Loops in Operation 1 e This page left blank pending NRC approval of three locp operation e e e 1 _ SEQUOYAH - UNIT 1 2-3

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor trip system instrumentation and interlocks setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY: As shown for each channel in Table 3.3-1. ACTION: With a reactor trip system instrumentation or interlock setpoint less conserva-tive than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. SEQUOYAH - UNIT 1 2-4 Amendment No. 12

i TABLE 2.2-1 (Continued) M i 2 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 3 + g [ FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES-21. Turbine Impulse Chamber Pressure - 5 10% Turbine Impulse i 11% Turbine Impulse (P-13) Input to Low Power Reactor Trips Pressure Equivalent Pressure Equivalent H Block P-7 i 22. Power Range Neutron Flux - (P-8) Low < 35% of RATED < 36% of RATED Reactor Coolant Loop Flow, and Reactor THERMAL POWER THERMAL POWER. Trip 23. Power Range Neutron Flux - (P-10) - > 10% of RATED > 9% of RATED Enable Block of Source, Intermediate, THERMAL POWER THERMAL POWER i and Power Range (low setpoint) Reactor Trips 24. Reactor Trip P-4 Not Applicable Not Applicable 25. Power Range Neutron Flux - (P-9) - < 50% of RATED < 51% of RATED Blocks Reactor Trip for Turbine THERMAL POWER THERMAL POWER Trip Below 50% Rated Power NOTATION I g l I + 5 I I NOTE 1: Overtemperature AT ( ) $ AT {K - K2 (1 + T2 )[T( )-T'] + K (P-P') - f)(AI)}- g j 3 1 + r)S 3 4 b I*I5 0 n )f = Lag compensator on measured AT y where: 1 g t) = Time constants utilized in the lag compensator for AT I3 l = 2 secs. AT = Indicated AT at RATED THERMAL POWER g K) 5 1.14 K = 0.009 2 9

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T Less Than or Equal to 200 F avn LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% delta k/k. C APPLICABILITY: MODE 5. ACTION: With the SHUTDOWN MARGIN less than 1.0% delta k/k, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% delta k/k: Within one hour after detection of an inoperable control rod (s) and a. at least once per 12 hours thereafter. while the' rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s), b. At least once per 24 hours by consideration of the following factors: 1. Reactor coolant system boron concentration, 2. Control rod position, 3. Reactor coolant system average temperature, 4. Fuel burnup based on gross thermal energy generation, 5. Xenon concentration, and 6. Samarium concentration. SEQUOYAH - UNIT 1-3/4 1-3 Amendment No.12 r __.

r REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT {' s-- LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MIC) shall be: Less positive than 0 delta k/k/ F for the all rods withdrawn, beginning a. of cycle life (BOL), hot zero THERMAL POWER condition. b. Less negative than -4.0 x 10'4 delta k/k/ f for the all rods with-drawn, end of cycle life (E0L), RATED THERMAL POWER condition. APPLICABILITY: Specification 3.1.1.3.a - MODES 1 and 2* only# Specification 3.1.1.3.b - MODES 1, 2 and 3 only# ACTION: With the MTC more positive than the limit of 3.1.1.3.a above operation a. in MODES 1 and 2 may proceed provided: 1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than 0 delta k/k/ f within 24 hours or be in HOT STANDBY within the next 6 hours. These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6. 2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MlC has been restored to within its limit for the all rods withdrawn condition. 3. In lieu of any other report required by Specification 6.9.1, a Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition. b. With the MTC more negative than the limit of 3.1.1.3.b above, be in HOT SHUTDOWN within 12 hours. SWith Keff greater than or equal to 1.0

See Special Test Exception 3.10.3 l

SEQUOYAH - UNIT 1 3/4 1-4 l e

REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN LIMITING CONDITION FOR OPERATION r 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE: a. A flow path from the boric acid tank via a boric acid transfer pump and charging pump to the Reactor Coolant System if only the boric acid storage tank in Specification 3.1.2.5a is OPERABLE, or b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if only the refueling water storage tank in Specification 3.1.2.5b is OPERABLE. APPLICABILITY: MODES 5 and 6. ACTION: With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE: At least once per 7 days by verifying that the temperature of the a. heat traced portion of the flow path is greater than or equal to 145 F when a flow path from the boric acid tanks is used. b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. SEQUOYAH - UNIT 1 3/4 1-7 ~~

n l,,,,

~.., '. -. ~

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING 1 ] LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE: 4 i a. The flow path from the boric acid tanks via a boric acid transfer j pump and a charging pump to the Reactor Coolant System. b. Two finw paths from the refueling water storage tank via charging pumps to the Reactor Coolant System. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k at 200 F within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE: At least once per 7 days by verifying that the temperature of the a. 4 heat traced portion of the flow path from the boric acid tanks is greater than or equal to 145*F when it is a required water source. b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. At least once per 18 months during shutdown by verifying that each I c. automatic valve in the flow path actuates to its correct position on i a safety injection test signal. d. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a delivers at least 10 gpm to the Reactor Coolant System. ( SEQUOYAH - UNIT 1 3/4 1-8 Amendment No. 12 0

3/4.2 POWER DfSTRfBUTION LIMITS g(.,,) 3/4.2.1 XIAL FLUX DIFFERENCE (AFD), ~ .e LIMITING CONDITION FOR OPERATION -a 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within a ~ 1 5% target band (flux difference units) about the target flux difference. APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER

ACTION:

With the indicated AXIAL FLUX DIFFERENCE outside of the 1 5% target a. band about the target flux difference and with THERMAL POWER: 1. Above 90% of RATED THERMAL POWER, within 15 minutes: a) Either restore the indicated AFD to within the target band limits, or b) Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER. 2. Between 50% and 90% of RATED THERMAL POWER: ' ~ a) POWER OPERATION may continue provided: 1) The indicated AFD has not been outside of the 1 5% target band for more than 1 hour penalty deviation cumulative during the previous 24 hours, and 2) The indicated AFD is within the limits shown on Figure 3.2-1. Otherwise, reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. b) Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification 4.3.1.1.1 provided the indicated AFD is maintained within the limits of Figure 3.2-1. A total of 16 hours operation may be accumulated with the AFD outside of the target band during this testing without penalty deviation. M See Special Test Exception 3.10.2. SEQUOYAH - UNIT 1 3/4 2-1 w. w,e e,,% e.-w..-.- -w_

POWER DISTRIBUTION LIMITS ACTION:-(Continued) b. ' THERMAL POWER shall not be increased above 90% of RATED THERMAL POWER unless the indicated AFD is within the + 5% target band and ACTION a.2.a)l), above has been satisfied. c. T11ERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD has not been outside of the + 5% target band for more than I hour penalty deviation cumulative during the previous 24 hours. Power increases above 50% of RATED THERMAL POWER do not require being within the target band provided the accumulative penalty deviation is not violated. SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by: Monitoring the indicated AFD for each OPERABLE excore channel: a. 1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2. At least once per hour for the first 24 hours after restoring the AFD Monitor Alarm to OPERABLE status. b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable. The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging. 4.2.1.2 The indicated AFD shall be considered cuiside of its + 5% target band when at least 2 OPERABLE excore channels are indi' < ting the AFD to be outside the target band. Penalty deviation outside of the : 5% target band shall be accumulated on a time basis of: a. One minute penalty deviation for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and b. One-half minute penalty deviation for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER. 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Ef fective Full Power Days. The provisions of Specification 4.0.4 are not applicable. ~SEQUOYAH - UNIT 1 3/4 2-2 Amendment No.12

~ POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) b) At least once per 31 EFPD, whichever occurs first. C RTP 2. When the F is less than or equal to the F limit for the xy xy ,l appropriate measured core plane, additional power distribution RTP maps shall be taken and F compared to F and F t least x x x once per 31 EFPD. e. The F limits for RATED THERMAL POWER (FRTP) shall be provided for xy xy all core planes containing bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.14. f. The F limits of e, above, are not applicable in the following core x plane regions as measured in percent of core height from the bottom of the fuel: y 1. Lower core region from 0 to 15%, inclusive. 2. Upper core region from 85 to 100%, inclusive. 3. Grid plane regions at 17.8 1 2%, 32.1 1 2%, 46.4 1 2%, 60.6 t 2% and 74.9 + 2%, inclusive. 4. Core plane regions within 1 2% of core height (1 2.88 inches) about the bank demand position of the bank "D" control rods. L g. With F exceeding Fxy, the effects of F on F (Z) shall be evaluated xy xy Q to determine if F (Z) is within its limit. q 4 f. i l r - SEQUOYAH - UNIT 1 3/4 2-7 Amendment No. 12 ,. ~ - _,.,. - -...., - - -

PC' DER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 1 v ' 4.2.2.3 When F (Z) is measured for other than F determinations, an overall q xy measured F (Z) shall be obtained from a power distribution map and increased g by 3% to account for manufacturing tolerances ar.d further increased by 5% to account for measurement uncertainty. F t SEQUOYAH - UNIT 1 3/4 2-8 e- -w-,m. - m ~ rw w - -m ,.W~*-Wm** ^

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POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOWRATE AND R LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R, R shall be maintained within the regions of allowable y 2 operation shown on Figure 3.2-3 for 4 loop operation: Where: F"H A R) 1.49 [1.0 + 0.2 (1.0 - P)] a. = b. R2 [1 - bP(Bu)] THERMAL POWER c* P = RATED THERMAL POWER

d.

F Me sured values of F obtained by using the movable = H g incore detectors to obtain a power distribution map. The measured values of F shall be used to calculate H R since Figure 3.2-3 includes measurement uncertainties N of 3.5% for flow and 4% for incore measurement of F nd AH, e. RBP (Bu) = Rod Bow Penalty as a function of region average burnup as shown in Figure 3.2-4, where a region is defined as those assemblies with the same loading date (reloads) or enrichment (first core). APPLICABILITY: MODE 1 ACTION: With the combination of RCS total flow rate and R), R utside the regions 2 of acceptable operation shown on Figure 3.2-3: a. Within 2 hours: 1. Either restore the combination of RCS total flow rate and R), R to within the above limits, or 2 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High trip setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. SEQUOYAH - UNIT 1 3/4 2-10 Amendment No.12 _-~

3/4.3 INSTRUMENTATION 3/4.3.l' REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and -interlocks of Table 3.3~1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2. APPLICABILITY: As shown in Table 3.3-1. ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTa 4.3.1.1.1 Each reactor trip system instrumentation channel and interlock shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencie's shown in Table 4.3-1. 4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceeding 92 days. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation. 4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ~ reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1. SEQUOYAH - UNIT 1 3/4 3-1 Amendment No. 12 g ~

~. TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION c s x MINIMUM e TOTAL NO. CHANNELS CHANNELS APPLICABLE } FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1. Manual Reactor Trip 2 1 '2 1, 2, and

1 2.

Power Range, Neutron Flux 4 2 3 1, 2 2 3. Power Range, Neutron Flux 4 2 3 1, 2 2 High Positive Rate 4. Power Range, Neutron Flux, 4 2 3 1, 2 2 High Negative Rate 5. Intermediate Range, Neutron Flux 2 1 2 1, 2, and

3 to 6.

Source Range, Neutron Flux ' gg 4 A. Startup 2 1 2 2 , and

4 B.

Shutdown 2 0 1 3, 4 and 5 5 7. Overtemperature Delta T Four Loop Operation 4 2 3 1, 2 6, g Three Loop Operation 4 1** 3 1, 2 9 k 8. Overpower Delta T Four Loop Operation 4 2 3 1, 2 6, P, Three Loop Operation 4 1** 3 1, 2 9 9. Pressurizer Pressure-Low 4 2 3 1, 2 6# 10. Pressurizer Pressure--High 4 2 3 1, 2 6 11. Pressurizer Water Level--High 3 2 2 1, 2 7 6

r k i TABLE 3.3-1 (Continued) S REACTOR TRIP SYSTEM INSTRUMENTATION g -E s MINIMUM e TOTAL NO. CHANNELS CHANNELS APPLICABLE { FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 19. Safety Injection Input from ESF 2 1 2 1, 2 12 20. Reactor Trip Breakers 2 1 2 1, 2, and

12 21.

Automatic Trip Logic 2 1 2 1, 2, and

12 22.

Reactor Trip System Interlocks A. Intermediate Range t' Neutron Flux P-6 2 1 2 2, and* Ba B. Power Range Neutron Y Flux - P-7 4 2 3 1 8b C. Power Range Neutron Flux - P-8 4 2 3 1 8c D. Power Range Neutron Flux - P-10 4 2 3 1, 2 Bd E. Turbine Impulse Chamber Pressure - P-13 2 1 2 1 8b p. g F. Power Range Neutron 8. Flux - P-9 4-2 3 1 8e R-G. Reactor Trip - P-4 2 1 2 1, 2, and* 14 5

1 INSTRUMENTATION TABLE 3.3-1 (Continued) TABLE NOTATION n With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal, and fuel in the reactor vessel. The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.

The provisions of Specification 3.0.4 are not applicable.
1. High voltage to detector may be de-energized above the P-6 (Block of Source Range Reactor Trip) setpoint.

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours and/or open the reactor trip breakers. ACTION 2.- With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and POWER OPERATION may proceed provided the following conditions are racisfied: The inoperable channel is placed 'n the tripped condition a. within 1 hour. b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1.1. Either, THERMAL POWER is restricted to less than or equal c. to 75% of RATED THERMAL and the Power Range, Neutron Flux high trip reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours. ~ d. The QUADRANT POWER TILT RATIO, as indicated by the remaining three detectors is verified consistent with the normalized symmetric power distribution.obtained by using the movable incore detectors in the four pairs of symmetric thimble locations.at least once per 12 hours when THERMAL POWER is greater than 75% of RATED THERMAL POWER. =SEQUOYAH - UNIT 1 3/4 3-5 Amendment No. 12

.INSTRUMENTAT10N 1 TABLE 3.3-1 (Continued) [. ACTION 3 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level: Below the P-6 (Block of Source Range Reactor Trip) setpoint, a. restore the inoperable channel to OPEP.ABLE status prior to increasing THERMAL POWER above the P-6 Setpoint. b. Above the P-6 (Block of Source Range Reactor Trip) setpoint, but below 5% of RATED THERFAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER i above 5% of RATED THERMAL POWER. Above 5% of RATED THERMAL POWER, POWER OPERATION may continue. c. d. Above 10% of RATED THERMAL POWER, the provisions of Specification 3.0.3 are not applicable. ACTION 4 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level: Below the P-6 (Block of Source Range Reactor Trip) setpoint, x a. restore the inoperable channel to OPERABLE status prior to i increasing THERMAL POWER above the P-6 Setpoint. b. Above the P-6 (Block of Source Range Reactor Trip) setpoint, operation may continue. ACTION 5 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUIDOWN MARGIN requirements of Specification 3.1.1.1 or.3.1.1.2, as applicable, within 1 hour and at least once per 12 hours thereafter. ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: The inoperable channel is placed in the tripped condition a. within 1 hour. b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1.1. ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed } until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within I hour. SEQUOYAH - UNIT 1 3/4 3-6 -\\ .~.

INSTRUMENTATION TABLE 3.3-1 (Continued) t ACTION 8 - With less than the Minimum Number of Channels OPERABLE, declare the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or apply the appro-priate ACTION statement (s) for these functions. Functions to be evaluated are: a. Source Range Reactor Trip ~ b. Reactor Trip Low Reactor Coolant Loop Flow (2 loops) Undervoltage Underfrequency Pressurizer Low Pressure Pressurizer High Level c. Reactor Trip Low Reactor Coolant Loop Flow (1 loop) d. Reactor Trip Intermediate Range Low Power Range i Source Range e. Reactor Trip i Turbine Trip ACTION 9 With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours or be in HOT STANDBY within the next 6 hours; however, one channel associated with an operating loop may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1.1. ACTION 10 - With one channel inoperable, restore the inoperable channel to OPERABLE status within 2 hours or reduce THERMAL POWER to below ~ the P-8 (Block Low Reactor Coolant Pump Flow) setpoint breaker within the next 2 hours. Operation below the P-8 (Block of Low Reactor Coolant Pump Flow) setpoint breaker may continue pursuant to ACTI0ii 11. ACTION 11 - With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour. ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to I hour for surveillance testing per Specification 4.3.1.1.1 provided the other channel is OPERABLE. SEQUOYAH - UNIT 1 3/43'7 Amendment No. 12 n - -,..., -, - -.

s n 4 (! } sa 7 i i TABLE 3.3-2 o REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES a 7 FUNCTIONAL UNIT RESPONSE TIME 1. Manual Reactor Trip NOT APPLICABLE 2. Power Range, Neutron Flux $ 0.5 seconds

l 1

3. Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE 4. Power Range, Neutron Flux, High Negative Rate 5 0.5 seconds

4 5.

Intermediate Range, Neutron Flux NOT APPLICABLE { 6. Source Range, Neutron Flux NOT APPLICABLE 4 j 7. Overtemperature Delta T $ 4.0 seconds

8.

Overpower Delta T NOT APPLICABLE-9. Pressurizer Pressure--Low $ 2.0 seconds' 10. Pressurizer Pressure--High 5 2.0 seconds 11. Pressurizer Water Level--High NOT APPLICABLE 12. Loss of Flow - Single Loop t (Above P-8) $ 1.0 seconds n Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel. e 8,

=, - -

8 a_ . t

j TABLE 3.3-2 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES w x e, FUNCTIONAL !! NIT RESPONSE TIME c 5 13. Loss. Flow _Two Loops [ (Above P-7 and below P-8) $ 1.0 seconds i 14. Main Steam Generator Water Level-- l Low-Low < 2.0 seconds 15. Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level NOT APPLICABLE 16. Undervoltage-Reactor Coolant Pumps 1 1.2 seconds 17. Underfrequency-Reactor Coolant Pumps 5 0.6 seconds 18. Turbine Trip w E A. Low Fluid Oil Pressure NOT APPLICABLE B. Turbine Stop Valve NOT APPLICABLE 19. Safety Injection Input from ESF NOT APPLICABLE \\ 20. Reactor Trip Breakers NOT APPLICABLE 21. Automatic Trip Logic NOT APPLICABLE k 22. Reactor Trip System Interlocks NOT APPLICABLE E. _,w

1-TABLE 4.3-1 M@ 1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Sg i CHANNEL MODES IN WHICH c CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE 1: } FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 1. Manual Reactor Trip N.A. N.A. S/U(1) 1, 2, and

t 2.

Power Range, Neutron Flux S D(2), M(3) M 1, 2 and Q(6) 3. Power Range, Neutron Flux, N.A. R(6) M 1, 2 } High Positive Rate i 4. Power Range, Neutron Flux, N.A. R(6) M 1, 2 - High Negative Rate 5. Intermediate Range, S R(6) S/U(1) 1, 2, and

w Neutron Flux 0

6. Source Range, Neutron Flux S(7) R(6) M and S/U(1) 2, 3, 4, 5, and

7.

Overtemperature Delta T S R M 1, 2 8. Overpower Delta T S R M 1, 2 E. 9. Pressurizer Pressure--Low S R H 1, 2 @g 10. Pressurizer Pressure--High 5 R M 1, 2 [ 11. Pressurizer Water Level--High S R M 1, 2 U 12. Loss of Flow - Single Loop S R M 1 13. Loss of Flow - Two Loops S R N.A. 1 8

I TABLE 4.3-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS I CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED -14. Main Steam Generator Water S R M 1, 2 Level--Low-Low 15. Steam /Feedwater Flow Mismatch and S R M 1, 2 Low Steam Generator Water Level 16. Undervoltage - Reactor Coolant N.A. R M 1 Pumps 17. Underfrequency - Reactor Coolant N.A. R M 1 Pumps 18. Turbine Trip A. Low Fluid Oil Pressure N.A. N.A. S/U(1) 1 w1 B. Turbine Stop Valve Closure N.A. N.A. S/U(1) 1 Y 19. Safety Injection Input from ESF N.A. N.A. M(4) 1, 2 20. Reactor Trip Breaker N.A. N.A. M(5) and S/U(1) 1, 2, and W 21. Automatic Trip Logic N.A. N.A. M(5) 1, 2, and

22.

Reactor Trip System Inte.-locks A. Intermediate Range N.A. R S/U(8) 2, and

Neutron Flux, P-6 B.

Power Range Neutron N.A. R S/U(8) 1 Flux, P-7 F C. Power Range Neutron N.A. R S/U(8) 1 g Flux, P-8 D. Power Range Neutron N.A. R S/U(8) 1, 2 l Flux, P-10 E. Turbine Impulse Chamber N.A. R S/U(8) 1 g Pressure, P-13 F. Power Range Neutron N.A. R S/U(8) 1 U Flux, P-9 G. Reactor Trip, P-4 N.A. R S/U(8) 1, 2, and

j

INSTRUMENTATION 7 TABLE 4.3-1 (Continued) NOTATION ~ With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal. (1) If not performed in previous 7 days. (2) Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent. l (3) - Compare incore to excore axial flux difference above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference greater than or equal to 3 percent. (4) Manual ESF functional input check every 18 months. (5) - Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS. (6) - Neutron detectors may be excluded from CHANNEL CALIBRATION. (7) - Below P-6 (Block of Source Range Reactor Trip) setpoint. (8) Logic only, each startup or when required with the reactor trip l system breakers closed and the control rod drive system capable of rod withdrawal if not performed in previous 92 days. i l i SEQUOYAH . UNIT 1 3/4 3-13 Amendment No. 12

j INSTRU"ENTAT10N 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION r~ LIMITING CONDITION FOR OPERATION A _. - s 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5. o APPLICABILITY: As shown in Table 3.3-3. ~ ACTION: a. With an ESFAS instrumentation channel or interlock trip setpoint less conservative than the value shown in the Allowable Values column of l Table 3.3-4, declare the channel inoperable and apply the applicable l ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value. b. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3. SURVEILLANCE RE00]REMENTS ~ f 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operati,ons for the MODES and at the frequencies shown in Table 4.3-2. n 4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation. 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS fun ~ shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3. SEQUOYAH - UNIT 1 3/4 3-14 (_ \\ g_m ..-_-m._.

/ t l TABLE 3.3-3 (Continued) M @S ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION i E MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE c. g FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 2. CONTAINMENT SPRAY a. Manual 2 1 2 1,2,3,4 20 'b. Automatic Actuation 2 1 2 1,2,3,4 15 Logic c. Containment Pressure-- 4 2 3 1,2,3 18 High-High 3. CONTAINMENT ISOLATION Y j a. Phase "A" Isolation g 1) Manual 2 1 2 1,2,3,4 20 w 2) From Safety Injection 2 1 2 1,2,3,4 15 Automatic Actuation Logic 4 b. Phase "B" Isolation l 1) Manual 2 1 2 1,2,3,4 20 2) Automatic 2 1 2 1, 2, 3, 4 15 Actuation Logic .1 3) Containment 4 2 3 1,2,3 18 t l Pressure-High-High I i 'f i r. J e-

9

'o

7____-______-_. a l TABLE 3.3-3 (Continued) m ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRilMENTATION 5

r 4

MINIMUM i TOTAL NO. CHANNELS CHANNELS APPLICABLE s-CUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION .m i c. Containment Ventilation Isolation 1) Manual 2 1 2 1,2,3,4 19 2) Automatic Isolation 2 1 2 1,2,3,4 15 Logic 3) Containment Gas 2 1 1 1,2,3,4 19 Monitor Radioactivity-High 4) Containment Purge 2 1 1 1,2,3,4 19 I Air Exhaust Monitor y Radioactivity-High i 5) Containment Particu-2 1 1 1,2,3,4 19 3 late Activity High oo 4. STEAM LINE ISOLATION a. Manual 1/ steam line 1/ steam line 1/ operating 1, 2, 3 25 steam line b. Automatic 2 1 2 1,2,3 23 k Actuation Logic i c. Containment Pressure-- 4 2 3 1,2,3 18 3 High-High d. Steam Flow in Two 1, 2, 3 2 ? Steam Lines--High 5 Four Loops 2/ steam line 1/ steam line 1/ steam line 16* s Operating any 2 steam lines Three Loops 2/ operating l /any 1/ operating 17 Operating steam line operating steam line steam line .C

,-~. ) -p, j TABLE 3.3-3 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION = N MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE jzi FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION,- c H COINCIDENT WITH EITHER 1,2,3 T --Low-Low 2T any 1T any 16* r avg Four Loops 1Tavg/ loop jgggvg 310$%E Operating in any 17 iT*80erating Three Loops 1T /oper-l T in Operating ati8691oop anyop879 ting two loop loop i OR, COINCIDENT WITH Steam Line Pressure-1,2,3 Low g Four Loops 1 pressure / 2 pressures 1 pressure 16* Operating loop any loops any 3 loops I Three Loops 1 pressure / 1### pressure 1 pressure in 17 Operating operating loop in any oper-any 2 oper-i ating loop ating loops I i 5. TURBINE TRIP & I FEE 0 WATER ISOLATION a. Steam Generator 3/ loop 2/ loop in 2/ loop in 1, 2, 3 16* Water Level-- any oper-each oper-High-High ating loop ating loop ( n.

i i j TABLE 3.3-3 (Continued) m i E g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 5 =c e MINIMUM c TOTAL NO. CHANNELS CHANNELS APPLICABLE { ( FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 6. AUXILIARY FEEDWATER j a. Manual. Initiation 2 1 2 1,2,3 24 b. Automatic Actuation 2 1 2 1,2,3 23 i Logic c. Main Stm. Gen. Water Level-Low-Low i. Start Motor Driven Pumps 3/stm. gen. 2/stm. gen. 2/stm. gen. 1,2,3 16 gg any stm gen. u> ii. Start Turbine-Driven Pump 3/stm. gen. 2/stm. gen. 2/stm. gen 1,2,3 16 any 2 stm. gen. d. S.I. Start Motor-Driven Pumps and Turbine '{ Driven Pump See 1 above (all S.I. initiating functions and requirements) 8. e. Station Blackout .g Start Motor-Driven P, Pump associated 2/ shutdown 1/ shutdown 2/ shutdown with the shutdown board board board 1, 2, 3 20 2 .o board and Turbine Driven Pump rw f. Trip of Main Feedwater Pumps Start Motor-Driven Pumps and Turbine Driven Pump 1/ pump 1/ pump 1/ pump 1, 2 20*- g. ' Auxiliary Feedwater Suction Pressure-Low 3/ pump 2/ pump 2/ pump 1, 2, 3 20*

t INSTRUMENTATION I TABLE 3.3-3 (Continued) t ACTION 21 - With the number of OPERABLE Channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed 2 provided the following conditions are satisfied: } The inoperable channel is placed in the tripped condition a. within 1 hour. 9 b. The Minimum Channels OPERABLE requirements is met; however, l one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1.1. } ACTION 22 - With less than the Minimum Number of Channels OPERABLE, declare the interlock inoperable and verify that' all affected channels of the functions listed below are OPERABLE or apply the appropriate ACTION statement (s) for those functions. Functions to be l evaluated are: a. Safety Injection i Pressurizer Pressure t b. Safety Injection High Steam Line Flow Steam Line Isolation High Steam Line Flow ] Steam Dump c. Turbine Trip Steam Generator Level High-High Feedwater Isolation Steam Generator Level High-High } ACTION 23 - With the number of OPERABLE channels one less than the Total i Number of Channels, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to 1 hour t for surveillance testing per Specification 4.3.2.1. r ACTION 24 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTOOWN within the j following 6 hours. ACTION 25 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or declare the associated valve inoperable and take the ACTION required by Specification 3.7.1.5. SEQUOYAH - UNIT 1 3/4 3-23 Amendment No. 12

j< l TABLE 3.3-4 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS a f FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E 1. SAFETY INJECTION, TURBINE TRIP AND Z FEEDWATER ISOLATION .I e a. Manual Initiation Not Applicable Not Applicable b. Automatic Actuation Logic Not Applicable Not Applicable i c. Containment Pressure--High 1 1.54 psig i 1.7 psig. -t d. Pressurizer Pressure--Low 1 1870 psig 1 1860 psig e. Dif ferential Pressure < 100 psi ~< 112 psi ) Between Steam Lines--High ~ Y f. Steam Flow in Two Steam Lines-- < A function defined as < A function defined as High Coincident with T --Low-Low Tollows: A Ap correspond-Tollows: A Ap corresponding avg or Steam Line Pressure--Low ing to 40% of full steam to 44% of full steam flow flow between 0% and 20% between 0% and 20% load and loed and then a op then a Ap increasing increasing linearly to a linearly to a op correspond-op corresponding to 110% ing to 111.5% of full steam of full steam flow at flow at full load full load T 2 540 F T,yg 1 538 F avg

)

1 600 psig steam line 1 580 psig steam line pressure pressure S. t '/ i n.

TABLE 3.3-4 (Continued) .Ny ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS s x ' FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES e 2. CONTAINMENT SPRAY a. Manuc1 Initiation .Not Applicable Not Applicable b. Automatic. Actuation Logic Not Applicable Not Applicable c. Containment Pressure--High-High 5 2.81 psig $ 2.97 psig 3. CONTAINMENT ISOLATION a. Phase "A" Isolation 1. Manual Not Applicable Not Applicable wh w 2. From Safety Injection Not Applicable Not Applicable A, Automatic-Actuation logic u b. Phase "B" Isolation 1. Manual Not Applicable Not Applicable 2. Autcmatic Actuation Logic Not Applicable Not Applicable 3. Containment Pressure--High-High $ 2.81 psig 5 2.97 psig c. Containment Ventilation Isolation 1. Manual Not Applicable Not Applicable 2. Automatic Isolation Logic Not Applicable Not Applicable

TABLE 3.3-4 (Continued) 1 vi j .G ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 85 FUNCTIONAL UNIT TRIP'SETPOINT A_L_LOWABLE VALUES -3 -3 sl 3. Containment Gas Monitor 1 8.5.x 10 pCi/cc 5 8.5 x 10 Ci/cc y Radioactivity-High -3 -3 4. Containment Purge Air Exhaust 5 8.5 x 10 Ci/cc 1 8.5 x 10 Ci/cc Monitor Radioactivity-high j -5 -5 5. Containment Particulate 5 1.5 x 10 Ci/cc $ 1.5 x 10 Ci/cc Activity-High jj 4. STEAM LINE ISOLATION i i a. Manual Not Applicable Not Applicable b. Automatic Actuation Logic Not Applicable Not Applicable us); c. . Containment Pressure--High-High 5 2.81 psig 1 2.97 psig d. Steam Flow in Two Steam lines-- < A function defined as < A function defined as 53 High Coincident with T --Low-Low follows: A Ap correspond-follows: A Ap corresponding li Or Steam Line PressureayEow ing to 40% of full steam to 44% of full steam flow flow between 0% and 20% between 0% and 20% load load and then a Ap and then a Ap increasing 1 increasing linearly to a linearly to a Ap correspond-Ap corresponding to ing to 111.5% of full steam 110% of full steam flow flow at full load at full load. T 1 540 F T 1 5 8"F 3yg avg 1 600 psig steam 1 580 psig steam line pressure line pressure 5. TURBINE TRIP AND FEEDWATER ISOLATION a. Steam Generator Water level-- < 75% of narrow range < 76% of narrow range High-High Instrument span each steam Instrument span each steam 4 generator generator C \\ (m~ ~.:. i

i TABLE 3.3-4 (Continued] ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5 7 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E 6. AUXILIARY FEEDWATER [ a. Manual Not Applicable Not Applicable b. Automatic Actuation Logic Not Applicable Not Applicable c. Main Steam Generator Water Level-low-low > 21% of narrow range > 20% of narrow range Instrument span each Instrument span each steam generator steam generator d. S.I. See 1 above (all SI Setpoints) e. Station Blackout 0 volts with a 5.0 second 0 volts with a 5.0 i 1.0 second g time delay time delay f. Trip of Main Feedwater N.A. N.A. "4 Pumps g. Auxiliary Feedwater Suction 1 2 psig (motor driven pump),1 1 psig (motor driven pump) Pressure-Low 2 6.5 psig (turbine driven 1 5.5 psig (turbine driven pump) pump) 7. LOSS OF POWER k a. 6.9 kv Shutdown Board Undervoltage a g-1. Loss of Voltage 0 volts with a 0 volts with a g 1.5 second time 1.5 1 0.5 second time delay delay

s

.o 2. Load Shedding 0 volts with a 0 volts with a 5.0-second time delay 5.0 i 1.0 second g N time delay 8. ENGINEERED SAFETY-FEATURE ACTUATION SYSTEM INTERLOCKS a. Pressurizer Pressure Manual Block of Safety Injection P-il 5 1970 psig 5 1980 psig s

i .s ~ ,t - e } TABLE 3.3-4 (Contir.ued) ,mj ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS s x FUNCTIONAL UNIT . TRIP SETPOINT ALLOWABLE VALUES i .E 8. ENGINEERED SAFETY FEATURE ACTUATION [ SYSTEM INTERLOCKS-(Continued) . 4 ' l T,yg b. . Prevents Manual Block of Safety Injection P-12 5 540 F $ 542 F c. Tavg Manual Block of Safety Injection, Steam Line Isolation, Block Steam Dump 2 540 F 2 538*F w3 d. Steam Generator Level Turbine Trip, Feedwater Isolation w h P-14 (See 5. above) 9. AUTOMATIC SWITCH 0VER TO CONTAINMENT SUMP a. RWST Level - Low 130" from tank base 130" 1 4" from tank' base COINCIDENT WITH -Containment-Sump Level - High 30" above elev. 680' 30" 2.5" above e' lev. 680' AND { Safety Injection (See 1 above for all Safety Injection Setpoints/ Allowable Valves) E 8 E .F. --s- .n.,,

r TdBLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 6. Steam Flow in Two Steam Lines-High Coincident with Steam Line-Pressure-Low a. Safety Injection (ECCS) $ 13.0(7)/23.0(1) b. Reactor Trip (from SI) < 3.0 c. Feedwater Isolation < 8.0(2) d. Containment Isolation-Phase "A"(3) }18.0(8)/28.0(9) e. Containment Ventilation Isolation Not Applicable f. Auxiliary Feedwater Pumps < 60 g. Essential Raw Cooling Water System 65.058)/75.0(9) h. Steam Line Isolation < 8.0 i. Emergency Gas Treatment System 38.0(9) 7. Containment Pressure--High-High a. Containment Spray 1 58.00(9) b. Containment Isolation-Phase "B" 1 65(8)/75(9) c. Steam Line Isolation < 7.0 ~ d. Containment Air Return Fan > 540.0 and i 660 i 1 8. Steam Generator Water Level--High-High Turbine Trip-Rear. tor Trip $ 2.5 a. b. Feedwater Isolation 5 11.0(2) 9. Main Steam Generator Water Level - Low-Low a.' Motor-driven Auxiliary < 60.0 Feedwater Pumps (4) b. Turbine-driven Auxiliary 1 60.0 Feedwater Pumps (5) t SEQUOYAH - UNIT 1 3/4 3-31 Amendment No. 12 9 +r,, .--r-w - -, -r-, - - -,,, ---~--n ,, -,., -y, r-+----

i TABLE 3.3-5 (Contir.ued) 1 -ENGINEERED SAFETY FEATURES RESPONSE TIMES ( s.s s. m INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS r 10. Station Blackout a a. ' Auxiliary Feedwater Pumps 5 60 11. Trip of Main Feedwater Pumos ~ a. Auxiliary Feedwater Pumps < 60 12. Loss of Power a. 6.9 kv Shutdown Board i $ 10 Undervoltage (Loss of Voltage) t 13. RWST Level-Low Coincident with Containment Sump Level-High and Safety Injection i a. Automatic Switchover to { Containment Sump < 250 1 14. Containment Purge Air Exhaust i Radioactivity - High Containment Ventilation Isolation _$ 10(6) a. .1 15. Containment Gas Monitor Radioactivity High Containment Ventilation Isolation _ 10(6) a. .16. Containment Particulate Activity High a. Containment Ventilation Isolation i 10(6) SEQUOYAH - UNIT 1 3/4 3-32 c; 9- "" N9ED ~ =

W e
"N+

" ^ - " - " -

e TABLE 4.3-2 (Continued) p y, ll is gg ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION I g SURVEILLANCE REQUIREMENTS =c e gj CHANNEL MODES IN WHICH

q CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 3.

CONTAINMENT ISOLATION a. Phase "A" Isolation

1) Manual N.A.

N.A. M(1) 1, 2, 3, 4

2) From Safety Injection N.A.

N.A. M(2) 1, 2, 3, 4 Automatic Actuation Logic b. Phase "B" Isolation 2

1) Manual N.A.

N.A. M(1) 1, 2, 3, 4 w f;

2) Automatic Actuation N.A.

N.A. M(2) 1, 2, 3, 4 Logic

3) Containment Pressure--

S R M 1,2,3 High-High k c. Containment Ventilation Isolation h-

1) Manual N.A.

N.A. M(1) 1, 2, 3, 4 E"

2) Automatic Isolation' Logic N.A.

N.A. M(2) 1, 2, 3, 4 55 a

3) Containment Gas Monitor 5

R M 1,2,3,4 0 Radioactivity-High 0

f F TABLE 4.3-2 (Continued) 8g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENIS = E CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

4) Containment Purge Air S

R M 1, 2, 3, 4 Exhaust Monitor Raoio-activity-High

5) Containment Particulate S

R M 1, 2, 3, 4 Activity-High 4. STEAM LINE ISOLATION w1 a. Manual N.A. N.A. M(1) 1, 2, 3 b. Automatic Actuation Logic N. A. N.A. M(2) 1, 2, 3 c. Containment Pressure-- S R M 1, 2, 3 High-liigh d. Steani Flow in Two Steam S R H 1,2,3 Lines--High Coincident with T -- L w-Low or Steam 1.ine avg Pressure--Low 5. TURBINE TRIP AND FEEDWATER ISOLATION a. Steam Generator Water 5 R M 1, 2, 3 Level--High-High 6. AUXILIARY FEEDWATER a. Manual N. A. N.A. M(1) 1, 2, 3 b. Automatic Actuation Logic N.A. N.A. M(2) i, 2, 3 05 \\

y + TABLE 4.3-2 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION y SURVEILLANCE REQUIREMENTS x CHANNEL MODES IN WHICH E CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE' Z FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED w c. Main Steam Generator Water S R M 1,2,3 -f Level-Low-Low I d. S.I. See 1 above (all SI surveillance requirements) e. Station Blackout N.A. R N.A. 1, 2, 3 f. Trip of Main Feedwater N.A. N.A. R 1, 2 Pumps g. Auxiliary Feedwater Suction N.A. R M 1,2,3 y 7. LOSS OF POWER w a. 6.9 kv Shutdown Board g Undervoltage 1. Loss of Voltage S R M 1, 2, 3, 4 2. Load Shedding S R N.A. 1,2,3,4 8. ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS a. Pressurizer Pressure, N.A. R (4) N.A. 1, 2, 3 P-11 >j b. T,yg, P-12 N.A. R (4) N.A. 1, 2, 3 c. . Steam Generator N.A. R (4) N.A. 1, 2 Level, P-14 e E 9. AUTOMATIC SWITCH 0VER TO CONTAINMENT SUMP a. RSWT Level Low S R M 1,2,3,4 COINCIDENT WITH Containment Sump Level - High S R M 1,2,3,4 AND Safety Injection (See 1 above for all Safety Injection Surveillance Requirements) O

i 1 I INSTRUMENTATION TABLE 4.3-2 (Continued) TABLE NOTATION (1) Manual actuation switches shall be-tested at least once per 18 months during shutdown. All other circuitry associated with manual safeguards actuation shall receive a CHANNEL FUNCTIONAL TEST at least once per 31 days. (2) Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS. J (3) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter. (4) The total interlock function shall oe demonstrated OPERABLE during CHANNEL CALIBRATION testing of each channel affected by interlock operation. i-1 1 } S h 3 t i i. \\ i SEQUOYAH - UNIT 1 3/4 3-38 Amendment No. 12 .i

TNSTRUMENTATION 3/4.3.3' MONITORING INSTRUMENTATION %v RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 9 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm / trip setpoints within the specified-limits. APPLICABILITY: As shown in Table 3.3-6. ACTION: With a radiation monitoring channel alarm / trip setpoint exceeding a. the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours or declare the channel inoperable. b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6. c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. k SURVEILLANCE REQUIREMEN'S 4.3.'3.l' Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-3. .f. '\\L SEQUOYAH - UNIT 1 3/4 3-39

l j TABLE 3.3-6 m 1 RADIATION MONITORING INSTRUMENTATION E MINIMUM CHANNELS APPLICABLE ALARM / TRIP. MEASUREMENT g INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION Z 1. AREA MONITORS g -l 4 a. Fuel Storage Pool Area 1 $ 15 mR/hr 10 - 10 mR/hr 26 b. Containment Area 1 1, 2, 3 and 4 N/A 1 - 107 R/hr*** 30 2. PROCESS MONITORS -3 7 a. Containment Purge Air 1 1, 2, 3, 4 & 6 5 8.5 x 10 Ci/cc 10 - 10 cpm 28 b. Containment { i. Gaseous Activity -3 7 i y a) Ventilation Isolation 1 ALL MODES 5 8.5 x 10 Ci/cc 10 - 10 cpm 28 g b)RCS Leakage Detection 1 1, 2, 3 & 4 N/A 10 - 10 cpm 27 i ii. Particulate Activity -5 7 a) Ventilation Isolation 1 ALL MODES 5 1.5 x 10 pCi/cc 10 - 10 cpm 28 b)RCS Leakage Detection 1 1, 2, 3 & 4 N/A 10 - 10 cpm '27 7 y c. Control Room Isolation 1 ALL MODES 5 400 cpm ** 10 - 10 cpm 29 E y d. Noble Gas Effluent Monitors 0 r, g "With fuel in the storage pool or building -5

- Equivalent to 1.0 x 10 Ci/cc
  - Measurement range by extrapolation 0

l ! -~ L l INSTRUMENTATION T BLE 3.3-6 (Continuedl TABLE NOTATION ACTION 26 - With the number of OPERABLE channels less than required b'y the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours. ACTION 27 - With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1. ACTION 28 - With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9. ACTION 29 - With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, within 1 hour initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation. ACTION 30 - With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, restore the inoperable Channel (s) to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours, and in at least HOT SHUTDOWN within the following 6 hours and in COLD SHUTDOWN within the subsequent 24 hours. ~ l SEQUOYAH - UNIT 1 3/4 3-41 Amendment No. 12

TABLE 4.3-3 v,. 5 gj RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5 ZC .h CHANNEL MOSES IN WHICH at CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE -4 INSTRUMENT ~ CHECK CALIBRAlION TEST REQUIRED w 1. AREA MONITORS a. Fuel Storage Pool Area S R M b. Containment' Area S R M 1, 2, 3 & 4 2. PROCESS MONITORS-a. Containment Purge Air Exhaust S R M 1, 2, 3, 4 & 6 I$ b. Containment i. Gaseous Activity i T a) Ventilation Isolation S R M ALL MODES $5 b) RCS Leakage Detection S R M 1, 2, 3, & 4 ii. Particulate Activity a) Ventilation Isolation S R M ALL MODES b) RCS Leakage Detectior. S R M 1, 2, 3, & 4 B' c. Control Room Isolation S R M ALL MODES E 8* d. Noble Gas. Effluent Monitors E .n f

With fuel in the storage pool or building U

6 [ D.,

= _ _ I e INSTRUMENTATION MOVABLE INCORE DETECTORS-LIMITING CONDITION FOR OPERATION i- 's -3.3.3.2~The movable incore detection system shall be OPERABLE with: l a. At least 75% of the detector thimbles, i b. A minimum of 2 detector thimbles per core quadrant, and a Sufficient movable detectors, drive, and readout equipment to map c. these thimbles. APPLICABILITY: When the movable incore detection system is used for: I Recalibration of the excore neutron flux detection system, a. b. Monitoring the QUADRANT POWER TILT RATIO, or i c. Measurement of F g, F (Z) and F q xy' ACTION: With the movable incore detection system inoperable, do not use the system for i the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. J. 4 i { SURVEILLANCE REQUIREMENTS { 4.3.3.2 The movable incore detection syste'm shall be demonstrated OPERABLE by normalizing eacn detector output when required for: 1 Recalibration of the excore neutron flux detection system, or a. j b. . Monitoring the QUADRANT POWER TILT RATIO, or c. Measurement of F H' I (Z) and F Q y. 4 g SEQUOYAH -. UNIT 1 3/4 3-43' Amendment No. 12 = : . ::- ~ ,r z = =. .=

IN5TRUMENTATION SEISMIC INSTPUMENTATION y LIMITING CONDITION FOR OPERATION -3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE. APPLICABILITY: At all times. .c ACTION: With one or more seismic monitoring instruments inoperable for more a. than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS ( 4.3.3.3.1 Each of the above seismic monitnring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-4. 4.3.3.3.2 Each of the above seismic monitoring instruments actuated during a seismic event shall be restored to OPERABLE status within 24 hours and a CHANNEL CALIBRATION performed within 5 days following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days i describing the magnitude, frequency spectrum and resultant effect upon facility features important to safety. f8 SEQUOYAH - UNIT 1 3/4 3-44 r

TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION 4 a MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE

1. Triaxial Time-History Accelerographs I

a. 0-XT-52-75A, Containment Elev. 734 0-1.0g 1

l

b. 0-XT-52-75B, Annulus Elev 680 0-1.0g 1*

c. 0-XR-52-77, Diesel Building Elev. 722 0-1.0g 1

2. Triaxial Peak Accelerographs

a. 0-XR-52-82, Auxiliary Building Elev.

689 0-5.0g 1

b. 0-XR-52-83, Auxiliary Building Elev.

736 0-5.0g 1

c. 0-XR-52-84, Control Building Elev.

732 0-5.0g 1

3. Biaxial Seismic Switches

a. 0-XS-52-79, Annulus Elev. 680 0.025-0.25g 1*

b. 0-XS-52-80, Annulus Elev. 680 0.025-0.25g 1*

c. 0-XS-52-81, Annulus Elev. 680 0.025-0.25g 1*

i

4. Triaxial Response-Spectrum Recorders

a. 0-XR-52-86, Annulus Elev. 680 2-25.4 Hz, 0.003-90g 1*

l

b. 0-XR-52-87, Reactor Containment 2-25.4 Hz, 0.003-90g 1

l Bldg. Elev. 734

c. 0-XR-52-88, Aux. CR Elev. 734 2-25.4 Hz, 0.003-90g 1

i

d. 0-XR-52-89, OB Bldg. 2A Elev. 713 2-25.4 Hz, 0.003-90g 1

^With reactor control room indication SEQUOYAH - UNIT 1 3/4 3-45 Amendment No.12 ' l

b TABLE 4.3-4 SEISMIC M0HITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST

1. Triaxial Time-History Accelerographs

a. 0-XT-52-75A, Containment Elev. 734 M*

R SA

b. 0-XT-52-758, Annulus Elev. 680**

M* R SA 1

c. 0-XR-52-77, Diesel Building Elev. 722 M*

R SA

2. Triaxial Peak Accelerographs

a. 0-XR-52-82, Auxiliary Building Elev. 689 NA R

NA

b. 0-XR-52-83, Auxiliary Building Elev. 736 NA R

NA

c. 0-XR-52-84, Control Building Elev. 732 NA R

NA

3. Biaxial Seismic Switches

a. 0-XS-52-79, Annulus Elev. 680**

M R SA 4

b. 0-XS-52-80, Annulus Elev. 680**

M R SA

c. 0-XS-52-81, Annulus Elev. 680**

M R SA

4. Triaxial Response-Spectrum Recorders

a. 0-XR-52-86**, Annulus Elev. 680 M

R NA 4

b. 0-XR-52-87, Reactor Containment NA R

NA Bldg. Elev. 734

c. 0-XR-52-88, Aux. CR Elev. 734 NA R

NA I

d. 0-XR-52-89, DB Bldg. 2A Elev. 713 NA R

NA

Except seismic trigger
- With reactor control room indications.

SEQUOYAH - UNIT 1 3/4 3-46 Amendment No. 12 1 f a v, p. p .-.m y r--.e

l l INSTRUMENTATION i l ACCIDENT MONITORING INSTRUMENTATION \\ LIMITING CONDITION FOR OPERATION i 3.3.3.7 The accident monitoring instrumentation channels shown in Table 3.3-10 I shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3. ACTION: With the number of OPERABLE accident monitoring instrumentation a. channels less than the Required Number of Channels shown in Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours. b. With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 48 hours or be in at least HOT SHUTOOWN within the next 12 hours. t i SURVEILLANCE REQUIREMENTS 4.3.3.7 Each accident monitoring instrumentation channel shall be demon-i strated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION { operations at the frequencies shown in Table 4.3-7. i ? i 1 t i i I SEQUDYAH - UNIT 1 3/4 3-55 Amendment No. 12

s 4 TABLE 3.3-10 mE g ACCIDENT MONITORING INSTRUMENTATION 5 7 MINIMUM e REQUIRED NO CHANNELS z -INSTRUMENT OF CHANNELS OPERABLE w w 1. Reactor Coolant THot (Wide Range) 2 1 1 2. Reactor Coolant TCold (Wide Range) 2 1 3. Containment Pressure 2 1 4. Refueling Water Storage Tank Level 2 1 l 5. Reactor Coolant Pressure 2 1 6. Pressurizer Level (Wide Range) 2 1 m s* 7. Steam Line Pressure 2/ steam line 1/ steam line E 8. Steam Generator Level - Wide 1/ steam generator 1/ steam generator 9. Steam Generator Level - Narrow 1/ steam generator 1/ steam generator 10. Auxiliary Feedwater Flow Rate 1/ pump 1/ pump [ 11. Reactor Coolant System Subcooling Margin Monitor 1 0 8 i g 12. Pressurizer PORV Position Indicator

2/ valve 1/ valve 1

m Pressuri er PORV Block Valve Position Indicator ** 2/ valve 1/ valve ,o 13. 3 E 14. Safety Valve Position Indicator 2/ valve 1/ valve i I 15. Containment Water Level (Wide Range) 2 1 16. In Core Thermocouples 4/ core quadrant 2/ core quadrant "Not applicable if the associated block valve is in the closed position.

- Not applicable if.the block valve is verified in the closed position with power to the valve operator removed.

O

TABLE 4.3-7 m- /p j ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS -5 g z CHANNEL CHANNEL e INSTRUMENT CHECK CALIBRATION 5 [ 1, Reactor Coolant THot (Wide Range) M R 2. Reactor Coolant TCold (Wide Range) M R -t 3. Containment Pressure M R 4. Refueling Water Storage Tank Level M R s 5. Reactor Coolant Pressure M R 6. Pressurizer Level M R w1 7. Steam Line Pressure M R 8. Steam Generator Level - Wide M R 9. Steam Generator Level - Narrow M R 10. Auxiliary Fcedwater Flowrate M R i m 11. Reactor Coolant System Subcooling M R { Margin Monitor m ) N 12. Pressurizer PORV Position Indicator M R 4 z 13. Pressurizer PORV Block Valve M R Position Indicator ~w 3 j 14. Safety Valve Position Indicator M R 1 l 15. Containment Water Level (Wide Range) M R 16. In Core Thermocouples M R .I.

flisTRU".ENTATION FfRE DETECTf0N fNSTRUMENTATf0N j LIMITING CONDITION FOR OPERATION 3.3.3.8 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-11 shall be OPERABLE. APPLICABILITY: Whenever equipment protected by the fire detection instrument is required to be OPERABLE.# ACTION: With the number of OPERABLE fire detection instrument (s) less than the minimum number OPERABLE requirement of Table 3.3-11: Within 1 hour establish a fire watch patrol to inspect the zone (s) a. with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect the containment at least once per 8 hours or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5. b. Restore the inoperable instrument (s) to OPERABLE status within 14 days or, in lieu of any other report required by Specifica-tion 6.9.1, prepare and submit a Special Report to the Commission pursuant to. Specification 6.9.2 within the next 30 days outlining I the action taken, the cause of the inoperability and the plans and i schedule for restoring the instrument (s) to OPERABLE status. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. c. SURVEILLANCE REQUIREMENTS 4.3.3.8.1 Fach of the above required fire detection instruments which are accessible during operation shall be demonstrated OPERABLE at least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST. Fire detection which are not accessible during plant operation shall be demonstrated OPERABLE by the performance of a CHANNEL FUNCTIONAL TEST during each COLD SHUTDOWN ~ exceeding 24 hours unless performed in the previous 6 months. 4.3.3.8.2 The NFPA Code 72D supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months. 4.3.3.8.3 The non-supervised circuits between the local fire protection panels and actuated equipment shall be demonstrated OPERABLE at least once per 6 months. The fire detection instruments located within the containment are not required to be OPERABLE during the performance of Type A Containment Leakage Rate Tests. 1 SEQUOYAH - UNIT 1 3/4 3-58 / ~_- ,w,.e -+.- - -- -v* ~

TABLE 3.3-11 I FIRE DETECTION INSTRUMENTS Q: Fire Minimum Instruments Operable Zone Instrument Location Ionization Photoelectric Thermal Infrared 235 Ctrl. Rod Dr. Eqpt. Rm. El. 759 4 236 Ctrl. Rod Dr. Eqpt. Rm. El. 759 4 237 Mech. Eqpt. Rm. El. 749 2 [ 238 Mech. Eqpt. Rm. El. 749 2 241 480-V XFMR Rm. lA El. 749 3 242 480-V XFMR Rm. lA El. 749 3 243 480-V XFMR Rm. 1B El. 749 3 ~ 244 480-V XFMR Rm. 1B El. 749 3 249 125-V Batt. Rm. I El. 749 1 250 125-V Batt. Rm. I El. 749 1 251 125-V Batt. Rm. II El. 749 1 252 125-V Batt. Rm. II El. 749 1 253 125-V Batt. Rm. III El. 749 1 254 125-V Batt. Rm. III El. 749 1 ~ 255 125-V Batt. Rm. IV El. /19 1 256 125-V Batt. Rm. IV El. 749 1 q 257 480-V Bd. Rm. 1B El. 749 4 258 480-V Bd. Rm. 1B El. 749 4 259 480-V Bd. Rm. lA El. 749 4 260 480-V Bd. Rm. lA El. 749 4 153 Add. Eqpt. Bldg. El. 740.5 4 155 Refuel Rm. El. 734 19 ~ 156 RB Access Rm. El. 734 2 ~ 157 RB Access Rm. El. 734 2 160 SG B1wdn. Rm. El. 734 4 8 I l SEQUOYAH - UNIT 1 3/4 3-59 a -_ _._ m

INSTRUMENTATION TABLE 3.3-11 (Continued) FIRE DETECTION INSTRUMENTS Fire Minimum Instruments Operable Zone Instrument Location Ionization Photoelectric Thermal Infrared '161 SG Blwdn. Rm. El. 734 4 162 EGTS Rm. El. 734 3 163 EGTS Rm. El. 734 3 164 EGTS Fltr. A El. 734 1 165 EGTS F1tr. A El. 734 1 166 EGTS F1tr. 8 El. 734 1 167 EGTS'Fltr. B El. 734 1 172 Mech. Eqpt. Rm. El. 734 1 173 Mech. Eqpt. Rm. El. 734 1 176 480-V Shtdn. Bd. Rm. lAl El. 734 2 188 480-V Shtdn. Bd. Rm. 2Al El. 734 2 177 480-V Shtdn. Bd. Rm. lAl El. 734 2 189 480-V Shtdn. Bd. Rm. 2Al El. 734 2 178 480-V Shtdn. BJ Rm. lA2 El. 734 2 190 480-V Shtdn. Bd. Rm. 2A2 El. 734 3 179 480-V Shtdn. Bd. Rm. lA2 El. 734 2 191 480-V Shtdn. Bd. Rm. 2A2 EL. 734 3 180 480-V Shtdn. Bd. Rm. 181 El. 734 2 -192 480-V Shtdn. Bd. Rm. 2B1 El. 734 2 ~ 181 480-V Shtdn. Bd. Rm. lB1 El. 734 2 193 480-V Shtdn. Bd. Rm. 281 El. 734 2 182 480-V Shtdn. Bd. Rm. 182 El. 734 3 194 480-V Shtdn. Bd. Rm. 2B2 El. 734 2 183 480-V Shtdn. Bd. Rm.182 El. 734 3 195 480-V Shtdn. Bd. Rm. 2B2 El. 734 2 184 6.9-KV Shtdn. Bd. Rm. A El. 734 6 i SEQUOYAH - UNIT 1 3/4.3-60 Amendment No. 12 e--- 77 y~~--. .m. 3.,, ,.w., y., ~. ,,,,,m m.-

TABLE 3.3-11 (Continued) FIRE DETECTION INSTRUMENTS 2 Fire Minimum Instruments Operable Zone Instrument Location Ionization Photoelectric Thermal Infrared 69 Mech. Equip. Rm. El. 669 2 70 Aux. Bldg. A5-All, Col. W-X, El. 669 5 71 Aux. Bldg. AS-All, Col. W-X, El. 669 5 72 Aux. FW Pump Turbine 1A-5, El. 669 1 73 Aux. FW Pump Turbine 1A-5, El. 669 1 76 S.I. & Charging Pump Rms. El. 669 5 77 S.I. Pump Rm. lA_, El. 669 1 78 S.I. Pump Rm. 1B, El. 669 1 79 Charging Pump Rm. IC, El. 669 1 80 Charging Pump Rm. 1B, El. 669 1 81 Charging Pump Rm. lA, El. 669 1 88 Aux. Bldg. Corridor Al-A8, El. 669 8 3 89 Aux. Bldg. Corridor Al-A8, El. 669 8 90 Aux. Bldg. Corridor A8-A15, El. 669 8 91 Aux. Bldg. Corridor A8-A15, El. 669 8 92 Aux. Bldg. Corridor Col. U-W, El. 669 4 93 Aux. Bldg. Corridor Col. U-W, El. 669 4 94 Valve Galley, El. 669 2 95 Valve Galley, El. 669 2 ~ 39 Cont. Spray Pump 1A-A, El. 653 2 i 40 Cont. Spray Pump 1B-B, El. 653 2 43 RHR Pump 1A-A, El. 653 2 1 -44 RHR Pump 18-8, El. 653 2 47 Aux. Bldg. Corridor, El. 653 10-SEQUOYAH - UNIT 1 3/4 3-65

Amendment _No.

12 t I

TABLE 3.3-11 (Continued) FIRE DETECTION INSTRUMENTS ) vl Fire Minimum Instruments Operable Zone Instrument location Ionization Photoelectric Thermal Infrared / 267 Aux. Instr. Rm. El. 685 8 268 Aux. Instr. Rm. El. 685 9 269 Computer Rm. El. 685 4 270 Computer Rm. El. 685 4 276 Intk. Pumping Sta. El. 690 & 670.5 15 354 Upr. Compt. Coolers, El. 778 4 352 Lwr. Compt. Coolers, El. 693 4 356 RCP 2, El. 693 2 357 RCP 2, El. 693 2 360 RCP 1, El. 693 2 361 RCP 1, El. 693 2 364 RCP 3, El. 693 2 365 RCP 3, El. 693 2 368 RCP 4, El. 693 g 2 369 RCP 4, El. 693 2 372 Reactor Bldg. Annulus 18 373 Reactor Bldg. Annulus 18 1 Diesel Gen. Rm. 28-B, El. 722 5 2 Diesel Gen. Rm. 28-B, El. 722 5 3 Diesel Gen. Rm. 1B-8, El. 722 5 4 Diesel Gen. Rm. 18-8, El. 722 5 5 Diesel Gen. Rm. 2A-A, El. 722 5 ~ 6 Diesel Gen. Rm. 2A-A, El. 722 5 7 Diesel Gen. Rm. lA-A, El. 722 5 8 Diesel Gen. Rm. lA-A, El. 722 5 SEQUOYAH - UNIT 1 3/4 3-66 E

,~ l INSTRUMENTATION i TABLE 4.3-8 (Continued) TABLE NOTATION i ^ During liquid additions to the tank. (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists: 1. Instrument indicates measured levels above the alarm / trip setpoint. 2. Circuit failure. (2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists: 1. Instrument indicates measured levels above the alarm setpoint. 2. Circuit failure. (3) The initial CHANNEL CALIBRATION shall be performed using one or more of the. reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have i been related to the initial calibration shall be used. (4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are made. i J SEQUOYAH - UNIT 1 3/4 3-73 Amendment No. 12 1..

.JfiSTRUMEf!TAQO3 !GD10ACTIVLGAS QULffRRERLHQ1GSMUMJfRiEfilblM i-J 3 /: LidillNLQ,QMlJIIQftf,0,1_Q PJ R ATION __ 4 3.3.3.10 The radioactive gaseous effluent monitoring instrumentation channels [.' j shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the ODCM. APPLICABILITY: As shown in Table 3.3-13 ACTION: I With a radioactive gaseous effluent monitoring instrumentation a. i channel alarm / trip setpoint less conservative than required by the above Specification, immediately suspend the release of radioactive i gaseous ef fluents monitored by the affected channel or declare the channel inoperable. b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in T.able 3.3-13. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. c; I 7 SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CAllBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9. 1 l 4 4 SEQUDYAH - UNIT 1 3/4 3-74 i-T " J _ _ _g_1__ f -

rL_1

.1

st c m =

TABLE 4.3-9 (Continued) y, Eg RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS s i CHANNEL MODES IN WHICH E CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE ') INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED "~ 5. AUXILIARY BUILDING VENTILATION SYSTEM

a. Noble Gas Activity Monitor D

M R(3) Q(2) l b. Iodine Sampler W N.A. N.A. N.A.

c. Particulate Sampler W

N.A. N.A. N.A.

d. Flow Rate Monitor 0

N.A. R Q

e. Sampler Flow Rate Monitor 0

N.A. R Q i 6. SERVICE BUILDING VENTILATION SYSTEM

a. Noble Gas Actvity Monitor D

M R(3) Q(2) ya

b. Flow Rate Monitor 0

N.A. R Q 5 b l s O t

INSTRUMENTATION TABLE 4.3-9 (Continued) TABLE NOTATION 4 At all times. During waste gas disposal system operation. During shield building exhaust system operation.

- - - During waste gas releases.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists: 1. Instrument indicates measured levels above the alarm / trip setpoint. 2. Circuit failure. (2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room i alarm annunciation occurs if any of the following conditions exists: { 1. Instrument indicates measured levels above the alarm setpoint. j 2. Circuit failure. I (3) The initial CHANNEL CALIBRATION shall be performed using one or more of j the reference standards certified by the National Bureau of Standards or i using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. (4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal: i t 1. One volume percent hydrogen, balance nitrogen, and . [ ~ 2. Four volume percent hydrogen, balance nitrogen. ( (5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal: l. One volume percent oxygen, balance nitrogen, and [ 2. Four volume percent oxygen, balance nitrogen. i r SEQUOYAH - UNIT 1 3/4 3-80 . Amendment No. 12 t

3/4.4 kE/4LiOR LOOLAN1 $Y51EM 3/4.4.1 REACER COOLANT LOOPS AND COOLANT CIRCULATION (, f STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All reactor coolant loops shall be in operation. APPLICABILITY: MODES 1 and 2.* ACTION: With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within I hour. SURVEILLANCE REQUIREMENT t 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours.

See Special Test Exception 3.10.4.

~ SEQUOYAH - UNIT 1 3/4 4-1 ..s ' :\\. - = *, .-m

L REACTOR COOLANT SYSTEM HOT STANDBY 4 i LIMITING CONDITION FOR OPERATION 1 3.4.1.2 a. At least two of the reactor coolant loops listed below shall be OPERABLE: 1. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump, 2. Reactor coolant Loop B and its associated steam generator and reactor coolant pump, 3. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, 4. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump. b. At least one of the above coolant loops shall be in operation.* APPLICABILITY: MODE 3 ACTION: With less than the above required reactor coolant loops OPERABLE, a. restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours, b. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation. SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least-the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 21 percent at least once per 12 hours. 4,4.1.2.3 At least one Reactor Coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. ^All reactor coolant pumps may be de-energized for up to 1 hour provided 4 (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature. SEQUOYAH - UNIT 1 3/4 4-la Amendment No. 12

REACTCF. COOLANT SYSTEM SURVEILLANCE' REQUIREMENTS (Continued)- (('- v' 4.4.3.2.4 In addition to the requirements of Specification 4.0.5 the repair welds and adjoining areas of the pressurizer relief line shall be examined, using improved UT detection and evaluation procedures which have been demon-strated to be, effective in detecting IGSCC, prior to entering MODE 4 whenever the plant has been in COLD SHUTDOWN for 72 hours or more 'I the examination has not been performed in the previous 6 months. In the event these 6-month period examinations find the piping free of unaccept-able indications for 3 successive inspections, the inspection interval shall e be extended to 36 month intervals ( 12 months to coincide with a scheduled refueling outage). In the event these 36-month period examinations find the piping free of unacceptable indications for 3 successive inspections, the inspection interval shall be extended to 80-month periods. ? SEQUOYAH - UNIT 1 3/4-4b i k 9

REACTOR COOLANT SYSTEM 1 SHUTDOWN LIMITING CONDITION FOR OPERATION 3 3.4.1.3 a. At least two of the reactor coolant and/or residual heat removal (RHR) loops listed below shall be OPERABLE: 1. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump, 2. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump, 3. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, 4. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump, 5. Residual Heat Removal Loop A, 6. Residual Heat Removal Loop B. l b. At least one of the above reactor coolant and/or RHR loops shall be in operation.** APPLICABILITY: MODE 4. ACTION: With less than the above required loops OPERABLE, immediately initiate a. corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours. l { b. With no reactor coolant or RHR loop in operation, suspend all opera-tions involving a reduction in baron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

- All reactor coolant pumps and residual heat removal pumps may be de-energized for up to 1 hour provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10 F below saturation temperature.

SEQUOYAH - UNIT 1 3/4 4-2 Amendment No. 12 m

= REACTOR COOLANT SYSTEM i SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s), if not in operation,.shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. i 4.4.1.3.2 The required steam generator (s) sha)1 be determined OPERABLE by verifying secondary side water level to be ' greater than or equal to 10 percent (wide range indication) at least once per 12 hours. 4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. i i d a 1 1 4 l 1 l SEQUOYAH - UNIT 1. 3/4 4-2a Amendment No. 12 i ~- = ...-.= ~ -. - - -.1

REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.4 Two# residual heat removal (RHR) loops shall be OPERABLE

and at least one RHR loop shall be in operation.**

APPLICABILITY: MODE 5. ACTION: With less than the above required RHR/ reactor coolant loops OPERABLE, a. immediately initiate corrective action to return the required RHR/ reactor coolant loops to OPERABLE status as soon as possible. b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. SURVEILLANCE REQUIREMENTS 4.4.1.4 The residual heat removal loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours.

0ne RHR loop may be inoperable for up to 2 huurs for surveillance testing provided the other RHR loop is OPERABLE and in operation.

Four filled reactor coolant loops with at least 2 steam generators having levels greater than or equal to 10 percent (wide-range indication) may be substituted for one RHR loop.

The normal or emergency power source may be inoperable.

4

- The RHR pumps may be de energized for up to I hour provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10 F below saturation temperature.

SEQUOYAH - UNIT 1 3/4 4-2b Amendment No. 12 e e-w

REACTOR COOLANT SYSTEM RELIEF VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.4.3.2 Two power relief valves (PORVs) and their associated block valves shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION: With one or more PORV(s) inoperable, within 1 hour either restore a. the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, b. With one or more block valve (s) inoperable, within I hour either restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. The provisions of Specification 3.0.4 are not applicable. c. SURVEILLANCE REQUIREMENTS 4.4.3.2.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by: Performance of a CHANNEL CALIBRATION, and a. b. by operating the valve through one complete cycle of full travel. 4.4.3.2.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by o'erating the valve through one complete cycle of full travel. 4.4.3.2.3 The emergency power supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by: Transferring motive and control power from the normal to the emergency a. power supply and - b. Operating the valves through a comflete cycle of full travel. SEQUOYAH - UNIT 1 3/4 4-4a Amendment No. 12 r-_

REACTOR C00LANT' SYSTEM 3/4.4.4 PRESSURIZER t LIMITING CONDITION FOR OPERATION i .i 3.4.4 The pressurizer shall be CPERABLE with a water volume of less than or equal to 1656 cubic feet (equivalent to an indicated level of less than or equal to 92% on narrow range instrumentation), and at least two groups of pressurizer heaters each having a capacity of at least 150 kw. 4 APPLICABILITY: MODES 1, 2 and 3. ACTION: With one group of pressurizer heaters inoperable, restore at least two a. groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours. 4 1 SURVEILLANCE REQUIREMENTS l 4.4.4.1 The preusurizer water volume shall be determined to be within its limit at least once per 12 hours. 4.4.4.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by measuring circuit current at least once per 92 days. I 4.4.4.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by transferring power from the normal to the emergency power supply and energizing the heaters. SEQUOYAH - UNIT 1 3/4 4-5 Amendment No. '12

I REACTOR COOLANT'SYS' TEM i l 3/4.4.5 STEAM GENERATORS c'- sw.s LIMITING CONDITION FOR OPERATION v 3.4.5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: 4 With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T bove 200*F. avg SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5. 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam i generator tube minimum sample size, inspection result classification, and the corresponding action r.equired shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators;-the tubes selected for these inspections shall be selected on a random basis except: Where experience in similar plants with similar water chemistry 4 a. indicates critical areas to be inspected, then at least 50% of the 4 tubes inspected shall be from these critical areas. .~ b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include: 1. All nonplugged tubes that previously had detectable wall penetra-tions (greater than 20%). 2. Tubes in those areas where experience has indicated potential problems. SEQUOYAH - UNIT 1 3/4 4-6 i 4 \\ = = =..

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE: The lower containment atmosphere particulate radioactivity monitoring a.

system, b.

The containment pocket sump level monitoring system, and The lower containment atmosphere gaseous radioactivity monitoring c. system. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only two of the above required leakage detection systems OPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours when the required gaseous or particulate radioactive monitoring system is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by: The lower containment atmosphere gaseous and particulate monitoring a. ~ i systems performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL ) FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and b. Containment pocket sump level monitoring system performance of CHANNEL CALIBRATION at least once per 18 months. i SEQUOYAH - UNIT.1 3/4 4-13 Amendment No. 12

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to: a. No PRESSURE BOUNDARY LEAKAGE, b. 1 GPM UNIDENTIFIED LEAKAGE, 1 GPM total primary-to-secondary leakage through all steam generators and 500 c. gallons per day through any one steam generator, d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and 40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System e. pressure of 2235 + 20 psig. f. 1 GPM leakage at a Reactor Coolant System pressure of 2235 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1. APPLICABILITY: MODES 1, 2, 3 and 4 ACTION: With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY a. within 6 hours and in COLD SHUTDOWN within the following 30 hours. b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at-least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. With any Reactor Coolant System Pressure Isolation Valve leakage c. greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 20 hours. SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by: 1 i SEQUOYAH - UNIT 1 3/4 4-14 Amendment No.12 __. E

REACTOR COOLANT SYSTEM. SURVEILLANCE REQUIREMENTS (Continued) a Monitoring the lower containment atmosphere particulate radioactivity a. monitor at least once per 12 hours. b. Monitoring the containment pocket sump inventory and discharge at least once per 12 hours. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump c. seals when the Reactor Coolant System pressure is 2235 + 20 psig at i least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into Mode 3 or 4. d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours. Monitoring the reactor head flange leakoff system at least once per e. 24 hours. 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing requirements required by Specification 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit: a. At least once per 18 months. b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours or more and if leakage testing has not been performed in the previous 9 months. Prior to returning the valve to service following maintenance, repair c. or replacement work on the valve. d. Within 24 hours following valve actuation due to automatic or manual action or flow through the valve. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4. t l SEQUOYAH - UNIT 1 3/4.4-15 Amendment No.12 ' j a i l

TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER ' FUNCTION 63-560 Accumulator Discharge ~ t 63-561 Accumulator Discharge 63-562 Accumulator Discharge 63-563 Accumulator Discharge 63-622 Accumulator Discharge 63-623 Accumulator Discharge 63-624 Accumulator Discharge 63-625 Accumulator Discharge 63-551 Safety Injection (Cold Leg) 63-553 Safety Injection (Cold Leg) 63-557 Safety Injection (Cold Leg) 1 63-555 Safety Inje'ction (Cold Leg) 63-632 Residual Heat Removal (Cold Leg) 63-633 Residual Heat Removal (Cold Leg) 63-634 Residual Heat Removal (Cold Leg) 63-635 Residual Heat Removal (Cold Leg) 63-641 Residual Heat Removal / Safety Injection (Hot Leg) 63-644 Residual Heat Removal / Safety Injection '(Hot Leg) 63-558 Safety Injection (Hot Leg) 63-559 Safety Injection (Hot Leg) 1 63-543 Safety Injection (Hot Leg) 63-545 Safety Injection (Hot Leg) 63-547 Safety Injection (Hot Leg) 63-549 Safety Injection (Hot Leg) 63-640 Residual Heat Removal (Hot Leg) 63-643 Residual Heat Removal (Hot Leg) 87-558 Upper Head Injection 87-599 Upper Head Injection 87-560 Upper Head Injection 87-561 Upper Head Injection 87-562 Upper Head Injection 87-563 Upper Head Injection FCV-74-1* Residual Heat Removal FCV-74-2* Residual Heat Removal ^These valves do not have to be leak tested following manual or automatic actuation or flow through the valve. SEQUOYAH - UNIT 1 3/4 4-15a Amendment No. 12

REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM i 4 LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with: A maximum heatup of 100 F in any one hour period. a. b. A maximum cooldown of 100*F in any one hour period. A maximum temperature change of less than or equal to 5 F in any one c. I hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves. APPLICABILITY: At all times. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours and reduce the RCS T and pressure to less avg than 200 F and 500 psig, respectively, within the following 30 hours. SURVEILLANCE REQUIREMENTS ~ 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine change in material properties, at the intervals required by 10 CFR 50, Appendix H in accordance with the schedule in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3. SEQUOYAH - UNIT 1 3/4 4-23 Amendment No. 12

TABLE 4.4-5 3g REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE 5= CAPSULE VESSEL LEAD' s NUMBER LOCATION FACTOR WITHDRAWAL TIME (EFPY) T 4 3.73 10 REFUELING U 140 3.73 3 X 220 3.73 5 Y 320 3.73 9 5 40 1.09 EOL V 176 1.09 STBY W 184 1.09 STBY w s [ Z 356 1.09 STBY a g> ifn G. o en rt

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3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) i 3/4.5.1 ACCUMULATORS COLD LEG INJECTION ACCUMULATORS ~ i ~ LIMITING CONDITION FOR OPERATION 3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with: a. The isolation valve open, b. A contained borated water volume of between 7857 and 8071 gallons of borated water, c. Between 1900 and 2100 ppm of boron, and d. A nitrogen cover pressure of between 385 and 447 psig. APPLICABILITY: MODES 1, 2 and 3.* ACTION: 4 With one cold leg injection accumulator inoperable, except as a result a. of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. b. With one cold leg injection accumulator inoperable due to the isola-tion valve being closed, either immediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.5.1.1.1 Each cold leg injection accumulator shall be demonstrated OPERABLE: a. At least once per 12 hours by: 1. Verifying, by the absence of alarms or by measurement of levels and pressures, the contained borated water volume and nitrogen cover pressure in the tanks, and 2. Verifying that each cold leg injection accumulator isolation valve is open. ^ Pressurizer pressure above 1000 psig. SEQUOYAH - UNIT 1 3/4 5-1 Amendment No. 12 l

EMEPGENCY CORE COOLING SYSTEMS (ECCS) SURVEILLANCE REQUIREMENTS (Continued) (:- d s .J ' b. At least once per 31 days and within 6 hours af ter each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the cold leg injection accumulator solution. At least once per 31 days when the RCS pressure is above 2000 psig c. by verifying that power to the isolation valve operator is disconnected + by removal of the breaker from the circuit. ~ I d. At least once per 18 months by verifying that each cold leg injection accumulator isolation valve opens automatically under each of the following conditions: i j 1. When an actual or a simulated RCS pressure signal exceeds the P-11 (Pressurizer Pressure Block of Safety Injection) setpoint, 2 2. Upon receipt of a safety injection test signal. i i 4.5.1.1.2 Each accumulator water level and pressure channel shall be demonstrated OPERABLE: At.least once per 31 days by the performance of a CHANNEL FUNCTIONAL a. TEST. [ b. At l_ east once per 18 months by the performance of a CHANNEL CALIBRATION. T b SEQUOYAH - UNIT l 3/4 5-2 ~~,! t ~ ~~ ~~

EMERGENCY CORE COULING SYSTEMS (ECCS)' 3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION i . 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with: ^ a. A contained borated water volume of between 370,000 and 375,000

gallons, i

l b. A boron concentration of between 2000 and 2100 ppm of boron, c. A minimum solution temperature of 60 F, and d. A maximum solution temperature of 105 F. APPLICABILITY: MODES 1, 2, 3 and 4. 4 ACTION: 1 With the RWST inoperable, restore the tank to OPERABLE status within 1 hour or be in at.least HOT STANDBY within 6 hours and in COLO SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE: a. At least once per 7 days by: 1. Verifying the contained borated water volume in the tank, and 2. Verifying the boron concentration of the water. b. At least once per 24 hours by verifying the RWST temperature. 4 SEQUOYAH - UNIT 1 3/4 5-13 Amendment No. 12

I 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: MODES 1, 2, 3 and 4. i ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. i SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAIhMENT INTEGRITY shall be demonstrated: At least once per 31 days by verifying that all penetrations

not a.

capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-2 of Specification 3.6.3. b. By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3. After each closing of each penetration subject to Type B testing, c. i except the containment air locks, if coened following a Type A or B test, by leak rate testing the seal with gas at P, 12 psig, and i verifying that when the measured leakage rate for t8ese seals is added to the leakage rates determined pursuant to Specification 4.6.1.2.d for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60 L ' a

Except valves, blind flanges, and deactivated automatic valves which are located inside the annulus or containment and are locked, sealed or otherwise secured in the closed position.

These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days. SEQUOYAH - UNIT 1 '3/4 6-1 Amendment No. 12 )

.CONTAffNENT SYSTEMS e-4 CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION I o., 3.6.1.2 Containment leakage rates shall be limited to: l An overall integrated leakage rate of less than or equal to L,, a. 0.25 percent by weight of the containment air per 24 hours at P, 12 Psi 9, b. A combined leakage rate of less than or equal to 0.60 L, for all penetrations and valves subject to Type B and C tests, when pressurized to P,. A combined bypass leakage rate of less than or equai to 0.25 L, for c. all penetrations ident'fied in Table 3.6-1 as secondary containment bypass leakage paths when pressurized to P,. APPLICABILITY: MODES 1, 2, 3 and 4. (' a ACTION: 1 With either (a) the measured overall integrated containment leakage rate exceeding 0.75 L,, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests execeding 0.6d L,, or (c) with the combined bypass leakage rate exceeding 0.25 L,, restore the overall integrated leakage rate to less than or equal to 0.75 L,, the combined leakage rate for all penetrations and valves subject to Type B and C tests to less than or equal to 0.60 L,, and the combined bypass leakage rate to less i than er equal to 0.25 L, prior to increasing the Reactor Coolant System temperature above zC0*F. i t 1 1 SEQUOYAH - UNIT 1 3/4 6-2 1 i t tM+ u% F ' gr m +,- ---m we m w- < w,,- g -m--

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CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION Z 3.6.1.3 Each containment air lock shall be OPERABLE with: a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and b. An overall air lock leakage rate of less than or equal to 0.05 L at a P, 12 psig. a APPLICABILITY: MODES 1, 2, o a..d 4. ACTION: With one containment air lock door inoperable: a. 1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours or lock the OPERABLE air lock door closed. 2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days. 3. Otherwise, be in at least HOT STANDBY within the next six hours and in COLD SHUTDOWN within the following 30 hours. 4. The provisions of Specification 3.0.4 are not applicable. b. With the containment air lock inoperable, except as the result of an inoperable air lock door.. maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours ~ or be in at least HOT STANDBY within the next six hours and in COLD SHUTDOWN within the following 30 hours. SEQUOYAH - UNIT 1 3/4.6-7 ' Amendment No. 12

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE: t After tech opening, except when the air lock is being used for ) a. multiple entries, then at least once per 72 hours, by verifying seal leakage less than or equal to 0.01 L, when the volume between the 4 door seals is pressurized to greater than or equal to 6 psig for at least 15 minutes, 1 b. By conducting an overall air lock leakage test at not less than P a (12 psig) and by verifying the overall air lock leakage rate is within its limit:# 1. At least once per six months, and 2. Prior to establishing CONTAINMENT INTEGRITY if opened when CONTAINMENT INTEGRITY was not required when maintenance has i been performed on the air lock that could affect the air lock sealing capability.* At least once per 6 months by verifying that only one door in each c. air lock can be opened at a time. s i

The provisions of Specification 4.0.2 are not applicable.

l

Exemption to Appendix "J" of 10 CFR 50.

SEQUOYAH - UNIT 1 3/4 6-8 Amendment No. 12 1 4

CONTAINMENT SYSTEMS / ' CONTAINMENT VENTILATION SYSTEM ~ P. s. LIMITING CONDI110N FOR OPERATION i. 1 3.6.1.9 One pair (one purge supply line and one purge exhaust line) of con-tainment purge system lines may be open; the containment purge supply and exhaust isolation valves in all other containment purge lines shall be closed. Operation with purge supply or exhaust isolation valves open for either purging .1 or venting shall be limited to less than or eaual to 1000 hours per 365 days. APPLICABILITY: H0 DES 1, 2, 3, and 4. ~ s ACTION: J With a purge supply or exhaust isolation valve open in excess of the above cumulative limit, or with more than one pair of containment pJrge system lines open, close the isolation valve (s) in the purge line(s) within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. J. ( SURVEILLANCE REQUIREMENTS 4.6.1.9.1 The position of the containment purge supply and exhaust isolation i valves shall be determined at least once per 31 days. 4.6.1.9.2 The cumulative time that the purge supply and exhaust isolation valves are open during the past 365 days shall be determined at least once per 7 days. 1 'SEQUOYAH - UNIT 1-3/4 6-15 AMENDMENT NO. 5 p i APR 1519a! ^

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent containment spray systems shall be OPERABLE with each L spray system capable of taking suction from the RWST and transferring suction to the containment sump. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one containment spray system inoperable, restore the inoperable spray l system to OPERABLE status within 72 hours or be in HOT STANDBY within the next 6 hours; restore the inoperable spray system to OPERABLE status within the l next 48 hours or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE: At least once per 31 days by verifying that each valve (manual, a. power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. b. By verifying, that on recirculatica flow, each pump develops a discharge pressure of greater than or equal to 140 psig when tested pursuant to Specification 4.0.5. At least once per 18 months during shutdown, by: c. 1. Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure--High-High test signal. 2. Verifying that each spray pump starts automatically on a Containment Pressure--High-High test signal. d. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed. l l 1 i i SEQUOYAH - UNIT 1 3/4 6-16 Amendment No. 12 I l r- ~

CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES 1 LIMITING CONDITION FOR OPERATION 3.6.3 The containment isolation valves specified in Table 3.6-2 shall be OPERABLE with isolation times as shown in Table 3.6-2. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one or more of the isolation valve (s) specified in Table 3.6-2 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either: Restore the inoperable valve (s) to OPERABLE status within 4 hours, a. or b. Isolate each affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position, or Isolate each affected penetration within 4 hours by use of at least c. -one closed manual valve or blind flange; or d. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.3.1 The isolation valves specified in Table 3.6-2 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair.or replacement work is performed on the valve or its associated actuator, control or power' circuit by performance of a cycling test and verification of isolation time. 4.6.3.2 Each isolation valve specified in Table 3.6-2 shall be demonstrated OPERABLE during the COLD SHUTDOWN'or REFUELING MODE at least once per 18 months by: Verifying that on a Phase A containment isolation test signal, each a. Phase A isolation valve actuates to its isolation position. b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position. SEQUOYAH - UNIT 1 3/4 6-17 Amendment No. 12 j" .+,_y - w w- ,,r- --..y-

m r CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) h-Verifying that on a Containment Ventilation isolation test signal, j -c. each Containment Ventilation Isolation valve actuates to its isolation position. 1 4 4.6.3.3 The isolation time of each power operated or automatic valve of ' Table 3.6-2 shall tua determined to be within its limit when tested pursuant to Specification 4.0.5. 4.6.3.4 Each Containment Purge isolation valve shall be demonstrated I OPERABLE within 24 hours after each closing of the valve, except when the valve is being used for multiple cyclings, then at least once per 72 hours, by verifying that when the measure leakage rate of these valves is added to the leakage rates determined pursuant to Specification 4.6.1.2.d for all + i other Type B and C penetrations, the combined leakage rate is < 0.60 L ' a i ) 4 f i ) i l. i SEQUOYAH~- UNIT 1 3/4 6-18 Amendment No. 12

r ~ ll TABLE 3.6-2 (Continued) f CONTAINMENT ISOLATION VALVES !? Ef VALVE NUMBER FUNCTION MAXIMUM ISOLATION TIME'(Seconds)' ~ Ei C. PHASE "A" CONTAINMENT VENT ISOLATION (Cont.) Z 13. FCV-30-50 Upper Compt' Purge Air Exh 4 14. FCV-30-51 Upper Compt Purge Air Exh 4 15. FCV-30-52 Upper Compt Purge Air Exh 4 16. FCV-30-53 Upper Compt Purge Air Exh 4 17. FCV-30-56 Lower Compt Purge Air Exh 4 18. FCV-30-57 Lower Compt Purge Air Exh 4 19. FCV-30-58 Inst Room Purge Air Exh 4 i 20. FCV-30-59 Inst Room Purge Air Exh 4 i 21. FCV-90-107 Cntmt Bldg LWR Compt Air Mon 5 22. FCV-90-108 Cntmt Bldg LWR Compt Air Mon 5 23. FCV-90-109 Cntmt Bldg LWR Compt Air Mon 5 24. FCV-90-110 Cntmt Bldg LWR Compt Air Mon 5 u, ]; 25. FCV-90-111 Cntmt Bldg LWR Compt Air Mon 5 i 26. FCV-90-113 Cntmt Bldg LWR Compt Air Mon 5 33 A3 27. FCV-90-114 Cntmt Bldg LWR Compt Air Mon 5 ) 28. FCV-90-115 Cntmt Bldg LWR Compt Air Mon 5 29. FCV-90-116 Cntmt Bldg LWR Compt Air Mon 5 30. FCV-90-Il7 Cntmt Bldg LWR Compt Air Mon 5 D. OTHER 1. FCV-30-46 Vacuum Relief Isolation Valve 25 2. FCV-30-47 Vacuum Relief Isolation Valve 25 3. FCV-30-48 Vacuum Relief Isolation Valve 25 5 4 4 9 n

.=. -- _ _ _ i l CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL [ HYDROGEN MONITORS { f LIMITING CONDITION FOR OPERATION U i 3.6.4.1-Two independent containment hydrogen analyzers shall be OPERABLE. l j APPLICABILITY: MODES 1 and 2. [ ACTION: With one hydrogen monitor inoperable, restore the inoperable monitor to OPERABLE f status within 30 days or be in at least HOT STANDBY within the next 6 hours. i t f i i SURVEILLANCE REQUIREMENTS t 1-l i q 4.6.4.1 Each hydrogen monitor shall be demonstrated OPERABLE by the performance r of a CHANNEL CHECK at least once per 12 hours, a CHANNEL FUNCTIONAL TEST at least once per 31 days, and at least once per 92 days on a STAGGERED TEST BASIS by l . performing a CHANNEL CALIBRATION using sample gas containing: t One volume percent hydrogen, balance nitrogen. f a. j .1 b. Four volume percent hydrogen, balance nitrogen. I L i' j. i s i t i i P ? 4 p SEQUOYAH - UNIT l' -3/4 6 Amendment No. 12-i

CONTAINMENT SYSTEMS CONTAINMENT AIR RETURN FANS LIMITING CONDITION FOR OPERATION ^ 1 3.6.5.6 Two independent containment. air return fans shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one containment air return fan inoperable, restore the inoperable fan to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.5.6 Each containment air return fan shall be demonstrated OPERABLE: At least once per 92 days on a STAGGERED TEST BASIS by: a. 1. Verifying that the fan motor current is 28 i 7.5 amps with the backdraft dampers closed, and 2. Verifying that with the fan off, the air return fan damper opens when a torque of less than or equal to 68.1 inch-pounds is appied to the counterweight. b. At least once per 18 months by verifying that the air return fan starts on an auto-start signal after a 10 1 minute delay and operates for at least 15 minutes. i ] f 'SEQUOYAH - UNIT 1 3/4 6-33 Amendment No.12 .~

CONTAlliMENT SYSTEMS FLOOR DRAINS fw LIMITING CONDITION FOR OPERATION 3.6.5.7 The ice condenser floor drains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the ice condenser floor drain inoperable, restore the flocr drain to OPEP.ABLE status prior to ircreasing the Reactor Coolant System temperature above 200 f. SURVEILLANCE REQUIREMENTS 4.6.5.7 Each ice condenser floor drain shall be demonstrated OPERABLE at least once per 18 months during shutdown by: Verifying that valve gate opening is not impaired by ice, frost or a.

debris,

( b. Verifying that the valve seat is not damaged, Verifying that the valve gate opens when a force of less than or c. equal to 48 lbs is applied, and d. Verifying that the drain line from the ice condenser floor to the containment lower compartment is unrestricted. (.. SEQUOYAH - UNIT 1 3/4 6-34 gh1

. PLANT SYSTEMS l AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION l 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with: Two motor-driven auxiliary feedwater pumps, each capable of being J a. powered from separate shutdown boards, and 1: b. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system. APPLICABILITY: MODES 1, 2 and 3. ACTION: With one auxiliary feedwater pump inoperable, restore the required a. auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDYBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours, b. With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours. With three auxiliary feedwater pumps inoperable, immediately initiate j' c. corrective action to restore at least one auxiliary feedwater pump j j to OPERABLE status as soon as possible. t j j SURVEILLANCE REQUIREMENTS 1 4.7.1.2 In addition to the requirements of Specification 4.0.5 each auxiliary. feedwater pump shall be demonstrated OPERABLE by : i a. Verifying that: 1. each motor-driven pump develops a differential pressure of greater than or equal to 1397 psid on recirculation flow. 2. the steam-turbine driven pump develops a differential pressure of greater than or equal to 1183 psid on recirculation flow when the secondary steam supply pressure is greater than 842 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3. e i SEQUOYAH - UNIT 1 3/4 7-5 Amendment No. 12 4 -.w w yw=.++. S . -- er s *-: - - ' r 7+ m :*P W +-, A* -**4 e.,s-m,r-e+W '*~n-v e"-

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PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3. each automatic control valve in the flow path is OPERABLE whenever the auxiliary feedwater system is placed in automatic control or when above 10% of RATED THERMAL POWER. b. At least once per 18 months during shutdown by: 1. Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an auxiliary feedwater actuation test signal and a low auxiliary feedwater pump suction pressure test signal. 2. Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of an auxiliary feedwater actuation test signal. At least once per 7 days by verifying that each non-automatic valve c. in the auxiliary feedwater system flowpath is in its correct position. SEQUOYAH - UNIT 1 3/4 7-6 Amendment No. 12

=. -. PLANT SYSTEMS t 3/4.7.2 -STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION ~ f 1 3.7.2 The temperatures of both the primary and secondary coolants in the steam generators shall be greater than 70 F when the pressure of either coolant .in the steam generator is greater than 200 psig. APPLICABILITY: At all times. . ACTION: With the requirements of the above specification not satisfied: Reduce the steam generator pressure of the applicable side to less a. than or equal to 200 psig within'30 minutes, and b. Perform an engineering evaluation to determine the effect of the ) overpressurization on the structural integrity of the steam generator. Determine that the steam generator remains acceptable for continued j operation prior to increasing its temperatures above 200 F. SURVEILLANCE REQUIREMENTS 4.7.2 The pressure in each side of the_ steam generator shall be determined I to be less than 200 psig at least once per-hour when the temperature of either .the primary or secondary coolant is less than 70 F. 1 i 1 i i i 1 SEQUOYAH - UNIT 1 3/4 7-11 Amendment No. 12 .m _,,

PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent component cooling water loops shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUIDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.3 At least two component cooling water loops shall be demonstrated OPERABLE: At least once per 31 days on a STAGGERED TEST BASIS by verifying that a. each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position. b. At least once per 18 months, during shutdown, by verifying that each component cooling system pump starts automatically on a Saf ety Injection test signal. SEQUOYAH - UNIT 1 3/4 7-12 Amendment No. 12

PLANT SYSTEMS-3/4.7.4 ESSENTIAL RAW COOLING WATER SYSTEM

LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent essential raw cooling water (ERCW) loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only one ERCW loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.4. At least two ERCW loops shall be demonstrated OPERABLE: At least once per 31 days by verifying that each valve (manual, a. power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position. b. At least once per 18 months, during shutdown, by: 1. Verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection test signal. 2. Verifying that each ERCW pump starts automatically on a Safety Injection test signal. t l SEQUOYAH - UNIT 1 3/4 7-13 Amendment No. 12 1

l PLANT SYSTEMS l' ' 3/4.7.5 ULTIMATE HEAT SINK l LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink shall be OPERABLE with: a. A minimum water level in the forebay at or above elevation 668 feet Mean Sea Level, USGS datum, and b. The average temperature of water at the ERCW system suction of less than or equal to 81 F. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the water level in the forebay less than 668 feet Mean Sea a. Level USGS datum be at least HOT STANDBY within 6 hours and in COLD SHUTOOWN within the following 30 hours. The forbay portable makeup water pumping station shall be deployed within 5 days. b. With the average temperature of the water at the ERCW system suction greater than 81 F be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIRMENTS 4.7.5.1 The ultimate heat sink shall be determined OPERABLE at least once per l i 24 hours by verifying the average temperature and water level to be within their limits. 4.7.5.2 The components of the makeup water system shall be. inspected and the forebay portable makeup pumps and drives tested at least once per192 days. ) I ) l I l SEQUOYAH - UNIT 1 3/4 7-14 ~ Amendment No. 12 I k m

l PLANT SYSTEMS 3/4.7.6 FLOOD PROTECTION LIMITING CONDITION FOR OPERATION 3.7.6 The flood protection plan shall be ready for implementation to maintain the plant in a safe condition. APPLICABILITY: When one or more of the following conditions exist: heavy rainfall conditions in the east Tennessee watershed, [ a. b. an early warning or alert that a critical combination of flood and/or high headwater levels may or have developed, an early warning or alert involving Fontana Dam, or c. d. recognizable seismic activity in the east Tennessee region. l ACTION: With a Stage I flood warning issued initiate and complete within a. 10 hours the Stage I flood protection procedure which shall include being in at least HOT STANDBY within 6 hours, with a SHUTDOWN MARGIN of at least 5% delta k/k and T less than or equal to 350 F within the following 4 hours. If wit 8Y8 10 hours following the issuance of a Stage I flood warning communications between the TVA Division of Water Resources and the Sequoyah Nuclear Plant cannot be verified, initiate and complete the Stage II flood protection procedure within the following 17 hours. With a Stage II flood warning issued initiate the Stage II flood protection plan in time to ensure completion before the predicted flooding of the site and no later than 17 hours prior to the predicted arrival time of the initial critical flood l level (697 ft ms1 winter and 703 ft msl summer). b. With a seismic event occurring after a critical combination of flood and/or lieadwater alerts are issued verify and maintain communications .2 between TVA Power Control Center and the Sequoyah Nuclear Plant within 6 hours or initiate and complete the Stage I flood protection plan within the following 10 hours. If communications have not been established upon completion of the Stage I flood protection plan initiate and complete the Stage II flood protection plan within the following 17 hours. With a Fontana Dam Alert issued verify and maintain communications c. between Fontana Dam and the Sequoyah Nuclear Plant with I hour or initiate and complete the Stage I flood protection plan within 10 hours. If communications have not been established upon comnletion of the Stage I flood protection plan initiate and complete the Stage II flood protection plan within the following 17 hours. SEQUOYAH - UNIT 1 3/4 7-15 Amendment No. 12

i PLANT SYSTEMS 3/4.7.6 FLOOD PROTECTION t . LIMITING CONDITION FOR OPERATION (Continued) i d. With either the Norris, Cherokee, Douglas, Fort Loudon, Fontana, Hiwassee, Apalachia, Blue Ridge or Tellico dam failed seismically, after a critical combination of flood and/or headwater alerts is } issued initiate and complete the Stage I flaod protection plan within 10 hours. Upon completion of the Stage I flood protection plan initiate and complete the Stage II flood protection plan within the following 17 hours. Both the Stage I and the Stage II ( flood protection plans will be terminated if it is determined that the potential for flooding the site does not exist. ~ i SURVEILLANCE REQUIREMENTS 4 4.7.6.1 The water level in the forebay shall be determined at least once per 8-hours when the water level is less than or equal to 697 feet Mean Sea Level USGS datum during October 1 through April 15, or 703 feet Mean Sea Level t USGS datum during April 16 through September 30; and at least once per 2 hours when the water level is above these limits. f 4.7.6.2 Communications between Sequoyah Nuclear Plant: and TVA Division of Water Resources shall be maintained every a. 3 hours during heavy rainfall condition in the east Tennessee t watershed. I b. and TVA Power Control Center shall be maintained every 3 hours I following a recognizable seismic event that has occurred when a critical combination of flood and/or headwater alert is issued. 1 Communications shall be maintained until it has been determined j that the potential for flooding the site does not exist. and Fontana Dam shall be maintained every hour when an alert c. involving Fontana Dam has been issued by TVA Division of Water Resources. i i i 4 P I SEQUOYAH - UNIT 1 3/4 7-16 Amendment No. 12

PLANT-SYSTEMS '3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 Two independent control room emergency ventilation systems shall be OPERABLE. APPLICABILITY: ALL MODES ACTION: MODES 1, 2, 3 and 4 With one control room emergency ventilation system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. MODES 5 and 6 With one control room emergency ventilation system inoperable, a. restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the control room emergency ventilation system in the recirculation mode. b. With both control room emergency air ventilation systems inoperable, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. The provisions of Specification 3.0.3 are not applicable in MODE 6. c. SURVEILLANCE REQUIREMENTS 4.7.7 Each control room emergency ventilation system shall be demonstrated OPERABLE: At least once per 12 hours by verifying that the control room a. air temperature is less than or equal to 104 F. 1 b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at j least 15 minutes. At least once per 18 months or (1) after any structural maintenance c. on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by: SEQUOYAH - UNIT 1 3/4 7-17 Amendment No.12 1 ..,-m. ,-4

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 1. Verifying that the cleanup system satisfies the in place testing acceptance criteria and uses the test procedures of Regulatory ~ Positions C.S.a, C.5.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI H510 Sections 8 and 9), and the system flow rate is 4000 cfm + 10%. 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory ~ Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978. 3. Verifying a system flow rate of 4000 cfm + 10% during system operation when tested in accordance with ANSI N510-1975. d. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representa-tive carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978. e. At least once per 18 months by: 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 4000 cfm + 10%. 2. Verifying that on a safety injection signal or a high radiation signal from the air intake stream, the system automatically diverts its inlet flow through the HEPA filters and charcoal adsorber banks. 3. Verifying that the system maintains the control room at a positive pressure of greater than or equal to 1/8 inch Water Gauge relative to the outside atmosphere at a system flow rate of 4000 cfm i 10% (3800 cfm recirculation and 200 cfm fresh air). ~ f. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm i 10%. g. Af ter each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm i 10%. SEQUOYAH - UNIT 1 3/4 7-18 Amendment No. 12 [

- PLANT SYSTEMS-- 3/4.7.8 AUXILIARY BUILDING Ga5 TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION ^

3. 7. 9 Two independent auxiliary building gas treatment filter trains.shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4. AClION: With one auxiliary building gas treatment filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN w~ thin the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.8 Each auxiliary building gas treatment filter train shall be demonstrated ~ OPERABLE:' i At least once per 31 days on a STAGGERED TEST BASIS by initiating, a. from the control room, flow through the HEPA filter and charcoal 'adsorber train and verifying that the system operates for at least 10 hours with the heaters on. 1 b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by: 1. Verifying that the cleanup system satisfies the in place testing acceptance criteria and uses the test procedures of l Regulatory Positions C.S.a, C.S.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 9000 cfm i 10%. 2. Verifying within 31 days 'after removal that a laboratory analysis of a representative carbon sample obtained in accordance with l Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978. 3. Verifying a system flow rate of 9000 cfm 1 10% during system operation when tested in accordance with ANSI N510-1975. i SEQUOYAH - UNIl 1 3/4 7-19 Amendment No. 12 s 1 2 4 .mi,,e mm, w,. e a -.e e --=---g

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~ PLANT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) ~ ' After every 720 hours of charcoal adsorber operation by verifying c. within 31 days after removal that a laboratory analysis of repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978. d. At least once per 18 months by: 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the filter train at a flow rate of l 9000 cfm i 10%. 2. Verifying that the filter trains start on a Containment ~ Phase A Isolation test signal; or a high radiation signal from the fuel pool radiation monitoring system or the auxiliary building ventilation monitoring system. 3. Verifying that the system maintains the spent fuel storage area and the ESF pump ~ rooms at a pressure equal to or more negative than minus 1/4 inch water gage relative the outside atmosphere while maintaining a vacuum relief flow greater than 2000 cfm and a total system flow of 9000 cfm 10%. 4. Verifying that the heaters dispite 32 1 3.2 kw when tested in accordance with ANSI N510-1975. After each complete or partial replacement of a HEPA filter bank by e. verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DDP when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flew rate of 9000 cfm i 10%. ] f. After each complete or partial replacement of a charcoal adsorber c bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm i 10%. u SEQUOYAH - UNIT 1 3/4 7-20 Amendment No. 12 I s

4 4 PLANT SYSTEMS 3/4.7.9 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.9. All safety-related snubbers shall be OPERABLE. The snubbers are shown in Tables 4.7.9.a and 4.7.9.b and are listed in Surveillance Instruction SNP SI-162. Any exemptions to the surveillance program are shown in Table 4.7.9.c and in SNP SI-162. APPLICABILITY: MODES 1, 2, 3 and 4. (MODES 5 and 6 for snubbers located on systems or partial systems required OPERABLE in those MODES.) ACTION: With one or more snubbers inoperable, within 72 hours replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system. r SURVEILLANCE REQUIREMENTS 4.7.9. Each safety-rela ted snubber shall be demonstrated OPERABLE by perf ormance of the following augmented inservice inspection program and the requirements of Specification 4.0.5. These snubbers are shown in Tables 4.7.9.a and 4.7.9.b, 1 and are listed in Surveillance Instruction SNP SI-162. Table 4.7.9.b is a detailed tabulation of the hydraulic snubbers which are also shown in Table 4.7.9.a. Any exemption to any portion of the surveillance program for any snubber is shown in Table 4.7.9.c. a. Inspection Groups The snubbers may be categorized into two major groups based on whether the snubbers are accessible or inaccessible during reactor operation. These major groups may be further subdivided into subgroups based on design, environment, or other features which may be expected to affect the OPERABILITY of the snubbers within the subgroup. Each subgroup or group may be inspected independently in accordance with 4.7.9.b through 4.7.9.h. b. Visual Inspection Schedule and Lot Size l The first inservice visual inspection of snubbers shall be completed by October 31, 1981, and shall include all snubbers on safety-related systems. If less than two (2) snubbers are found inoperable during the first inservice visual inspection, the second inservice visual inspection shall be performed 12 months 25% from the date of the first inspection. Otherwise, subsequent visual inspections shall be performed in accordance with the following schedule: SEQUOYAH - UNIT 1 3/4 7-21 Amendment No. 12

PLANT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) b. Visual Inspection Schedule and Lot Size (Continued) Number of Inoperable Snubbers per Subsequent Visual Inspection Period Inspection Period *g 0 18 months i 25% 1 12 months 25% 2 6 months 25% 3, 4 124 days 25% 5,6,7 62 days 25% 8 or more 31 days i 25% c. Visual Inspection Performance and Evaluation Visual inspections shall verify (1) that there are no visible indications of damage or impaired OPERABILTY, (2) bolts attaching the snubber to the foundation or supporting structure are secure, and (3) snubbers attached to sections of safety-related systems that have experienced unexpected potentially damaging transients since the last inspection period shall be evaluated for the possibility of concealed damage and functionally tested, if applicable, to confirm operability. 1 Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible; and (2) the affected snubber is functionally tested, if applicable, in the as-found condition and determined OPERABLE per Specification 4.7.9.e. Hydraulic snubbers with inoperable single or common fluid reservoirs which have uncovered fluid ports shall be declared inoperable. When hydraulic snubbers which have uncovered fluid ports are tested, the tests shall be performed by starting with the piston at the as-found setting and extending the piston rod in the extension mode direction. Also, snubbers which have been made inoperable as the result of unexpected transients, isolated damage or other such random events, when the provisions of 4.7.9.g and 4.7.9.h have been met and any other appropriate corrective action implemented, shall not be counted in determining the next visual inspection interval. d. Functional Test Schedule, Lot Size, and Composition During each refueling outage, a representative sample of 10% of the total of the safety-related snubbers in use in the plant shall be functionally tested either in place or in a bench test. ^ The inspection interval shall not be lengthened more than one step at a time.

The provisions of Specification 4.0.2 are not applicable.

j SEQUOYAH - UNIT 1 3/4 7-22 Amendment No. 12

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) d. Functional Test Schedule, Lot Size, and Composition (Continued) ' The representative sample selected for functional testing shall include the various configurations, operating environments, and the range of size and capacity of snubbers within the groups or subgroups. The representative sample should be weighted to include more snubbers from severe service areas such as near heavy equipment. Unless a failure analysis as required by 4.7.9.1 indicates otherwise, the sample shall be a composite based on the ratio of each group to the total number of snubbers installed in the plant. Snubbers placed in the same location as snubbers which failed the previous functional test shall be included in the next test lot if the failure analysis shows that failure was due to location. ~ l The security of fasteners for attachment of the snubbers to the component and to the snubber anchorage shall be verified on snubbers selected for functional tests. { Functional Test Acceptance Criteria e. The ' snubber functional test shall verify that: 1. Activation (restraining action) is achieved within the specified range in both tension and compression, except that inertia dependent, acceleration limiting mechanical snubbers, may be tested to verify only that activation takes place in both directions of travel. 2. Snubber bleed, or release where required, is present in both tension and compression, within the specified range. 3. The force required to initiate or maintain motion of the snubber is within the specified range in both directions of travel. Also, the increase in the force required shall not exceed 50 percent of the 4 amount required at the last surveillance test of that snubber, provided that the force required is at least 25 pounds. 4. For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displace-ment shall be verified. 5. Testing methods may be used to measure parameters indirectly or parameters other than those specified if those results can be correlated to the specified parameters through established methods, f. Functional Test Failure Analysis and Additional Test Lots If any snubber selected for functional testing either fails to lock up or fails to move due to manufacture or design deficiency, all snubbers of the same design subject to the same defect shall be functionally tested. SEQUOYAH - UNIT 1 3/4 7-23 Amendment No.12

PLANT SYSTEMS l SURVEILLANCE REOUIREMENTS (Continued) f. Functional Test Failure Analysis and Additional Test Lots (Continued) If more than two snubbers do not meet the functional test acceptance criteria, an additional lot equal to one-half the original lot size shall be functionally tested for each failed snubber in excess of the two allowed failures. An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to determine the cause of the failure. The rasult of this analysis shall be used, if applicable, in selecting snubbers to be tested in the subsequent lot in an effort to determine the operability of other snubbers which may be subject to the same failure mode. (Selection of snubbers for future testing may also be based on the failure analysis.) Testing shall continue until not more than one additional inoperable snubber is found within a subsequent required lot or all snubbers of the original inspec-tion group have been tested, or all suspect snubbers identified by the failure analysis have been tested, as applicable. The discovery of loose or missing attachment fasteners will be evaluated to determine whether the cause may be localized or generic. The result of the evaluation will be used to select other suspect snubbers for verifying the attachment fasteners, as applicable. Snubbers shall not be subjected to prior maintenance specifically for the purpose of meeting functional test requirements. g. Functional Test Failure - Attached Component Analysis For snubber (s) found inoperable, an engineering evaluation shall be performed on the components which are restrained by the snubber (s). The purpose of this engineering evaluation shall be to determine if the components restrained by the snubber (s) were adversely affected by the inoperability of the snubber (s), and in order to ensure that the restrained component remains capable of meeting the designed service. h. Functional Testing of Repaired and Spare Snubbers Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meat the functional test criteria before installation in the unit. These snubbers shall have met the acceptance criteria subsequent to their most recent service, and the functional test must have been performed within 12 months before being installed in the unit. i. Snubber Service Life Program The seal service life of hydraulic snubbers shall be monitored to ensure that the seals do not fail between surveillance inspections. The maximum ) expected service life for the various seals, seal materials, and applica-tions shall be estimated based on engineering information, and the seals shall be replaced so that the maximum expected service life does not SEQUOYAH - UNIT 1 3/4 7-24 Amendment No. 12 l 1 - - - - - - - ~~ -

I PLANT-SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) t I i. Snubber Service Life Program (Continued) expire during a period when the snubber is required to be operable. The I seal replacements shall be documented and the documentation shall be retained in accordance with 6.10.2.n. Mechanical snubber drag force increases greater than 50 percent of previously measured values shall be e j evaluated as an indication of impending failure of the snubber. These evaluations and any associated corrective action shall be documented and .the documentation shall be retained in accordance with 6.10.2.n. j. Exemption From Visual Inspection or Functional Tests Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and if applicable snubber life destructive testing l was performed to qualify snubber operability for the applicable design conditions at either the completion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in Table 4.7.9.c and shall continue to be listed in the plant instruction SNP SI-162 indicating the extent of the exemptions. i i t b 4 } SEQUOYAH - UNIT 1 3/4 7-25 Amendment No. 12 i . ~. .~...4,

PLANT SYSTEMS TABLE 4.7.9a SAFETY-RELATED SNUBBERS

ACCESSIBLE INACCESSIBLE Small Medium & Large Hyd.

Small Medium & Large Paul Hyd. PSA PSA PSA PSA Munroe Size 1/4 1/2 1 3 10 35 100 1/4 1/2 1 3 10 35 100 MS 22 9 3 9 9 7 12 1 20 16 AMS 1 2 3 AFD 4 1 1 4 5 1 2 2 2 FD 2 6 1 1 2 1 CC 8 10 8 10 21 5 4 2 4 SI 3 1 2 37 12 2 9 15 1 4 CS 3 4 2 3 1 15 1 CVC 7 .5 1 4 24 7 3 8 1 RC-15 16 29 40 19 8 UHI 1 4 7 20 24 5 1 SGB 1 1 1 7 8 5 FPC 2 4 3 1 ERCW 2 5 4 22 19 23 15 RHR 5 2 2 6 2 2 1 3 IC 8 6 5 WD 9 DW 1 1 1 SA 1 1 1 PW 1 2 1 1 AC&H 12 7 Sub Total 44 24 19 31 30 11 7 154 86 87 106 77 15 Totals 68 98 47 240 285 20 31

Snubbers may be added to safety related systems without prior License Amendment to Table 4.7.9a provided that a. revision to Table 4.7.9a is included with the next License Amendment request.

Any exemptions t'o the provisions of the surveillance program for any snubber is indicated in Table 3.7.9c. SEQUOYAH - UNIT 1 3/4 7-26 Amendment No. 12

i s F TABLE 4.7.9b Nj SAFETY RELATED HYDRAULIC SNUBBERS

ACCESSIBLE INACCESSIBLE c

Sub Sub Size 1 1 2 2 3% 4 5 6 8 Total Size 1 1 2 2 3h 4 5 6 8 Total MS 1 5 5 1 12 MS 8 8 16 AMS 3 3 AMS 0 l AFD 2 3 5 AFD 2 2 FD 3 2 1 6 FD 1 1 CC 1 2 3 3 1 10 CC 2 2 4 SI 1 1 2 SI 1 2 1 4 CS 1 1 1 3 CS 0 ws* CVC 4 4 CVC 0 w 4 RC 0 RC 0 UHI 0 UHI 1 1 SGB 0 SGB 0 FPC 0 FPC 0 RHR 2 2 RHR 2 1 3 IC 0 IC 0 WD 0 WD 0 M DW 0 DW 0 m SA 0 SA 0 to 5 PW 0 PW 0 E AC&H 0 AC&H 0 ERCW 0 ERCW 0 -w Total 47 Total 31

Snubbers may be added to safety related systems without prior License Amendment to Table 4.7.9b provided that a revision to Table 4.7.9b is included with the next License Amendment request.

Any exemptions to the provisions of the surveillance program for any snubber is indicated in Table 3.7.9.c.

l PLANT SYSTEMS TABLE 4.7.9c SAFETY RELATED SNUBBERS - EXEMPTIONS TO THE SURVEILLANCE PROGRAM (EXEMPTED SNUBBERS TO BE ADDED LATER.) SEQUOYAH - UNIT 1 3/4 7-28 Amendment No.12 w y 9 g p =+4 9 7

PLANT SYSTEMS 3/4.7.10 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.10 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 micro-curies of removable contamination. APPLICABILITY: At all times. ACTION: a. With a sealed source having removable contamination in excess of the above limits, immediately withdraw the sealed source from use and: 1. Either decontaminate and repair the sealed source, or 2. Dispose of the sealed source in accordance with Commission Regulations. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE RE0VIREMENTS 4.7.10.1 Test Requirements - Each sealed source shall be tested for leakage -and/or contamination by: a. The licensee, or b. Other persons specifically authorized by the Commission or an Agreement State. The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample. 4.7.10.2 Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below. a. Sources in use - At least once per six months for all sealed sources containing radioactive materials: l. With a half-life greater than 30 days (excluding Hydrogen 3), and { 2. In any form other than gas. SEQUOYAH - UNIT 1-3/4 7-29 Amendment No. 12 t -.. - ~..

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous six months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use. c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source. 4.7.10.3 Repor's - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable contamination. -:SEQUOYAH - UNIT 1 3/4 7-30 Amendment No. 12 s

PLANT SYSTEMS 4 3/4.7.11 FIRE SUPPRESSION SYSTEMS i FIRE SUPPRESSION WATER SYSTEM 4 LIMITING CONDITION FOR OPERATION 1 L 3.7.11.1 The fire suppression water system shall be OPERABLE with: 2 Two fire suppression pumps, each with a capacity of 1174 gpm, with a. their discharge aligned to the fire suppression header, b. An OPERABLE flow path capable of taking suction from the forebay i and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water pressure alarm device on each sprinkler or hose standpipe, and the last valve ahead of the deluge valve on each deluge or spray system required to be OPERABLE i per Specifications 3.7.11.2 and 3.7.11.4. t { APPLICABILITY: At all times. ACTION: i I With only one pump OPERABLE, restore the inoperable equipment to l a. OPERABLE status within 7 days or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the plans and procedures to be used to restore the inoperable equipment to OPERABLE status or to provide an alternate backup pump or t supply. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. I b. With the fire suppression water system otherwise inoperable: i 1. Establish a backup fire suppression water system within 24 I hours, and 2. In lieu of any other report required by Specification 6.9.1, submit a Special Report in accordance with Specification 6.9.2: a) By telephone within 24 hours, b) Conci: ield by telegraph, mailgram or facsimile transmission no later than the first working day following the event, and I SEQUOYAH - UNIT 1 3/4 7-31 Amendment No.12

l PLANT SYSTEMS ACTION: (Continued) c) In writing within 14 days following the event, outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status. SURVEILLANCE REQUIREMENTS i 4.7.11.1 The fire suppression water system shall be demonstrated OPERABLE: At least once per 31 days on a STAGGERED TEST BASIS by starting each a. electric motor driven pump and operating it for at least 15 minutes on recirculation flow. b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path is in its correct position. At least once per 6 months by performance of a system flush. c. d. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel. At least once per 18 months by performing a system functional test e. which includes simulated automatic actuation of the system throughout J its operating sequence, and: 1. Verifying that each automatic valve in the flow path actuates to its correct position, 2. Verifying that each pump develops at least 1174 gpm at a system head of 312 feet, 3. Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and 4. Verifying that the No.1 fire pump starts to maintain the fire suppression water system pressure greater than or equal to 125 psig and that the No. 2 fire pump also starts automatically within 10 + 2 seconds when the fire suppression water system is not maintained greater than or equal to 125 psig by the No.1 pump. f. At least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association. SEQUOYAH - UNIT 1 3/4 7-32 Amendment No.12 I

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=_ I f PLANT SYSTEMf SPRAY AND/0R SPRINKLER SYSTEMS. t LIMITING CONDITION FOR OPERATION i i .i i 3.7.11.2 The following spray and/or sprinkler systems shall be OPERABLE: { i 1 a. Reactor Building - RC pump area, Annulus t b. Auxiliary Building - Elev. 669, 690, 706, 714, 734, 749, 759, ABGTS i Filters, EGTS Filters, Cont. Purge Filters, I and 125V Battery Rooms. [ a c. Control Building - Elev. 669, Cable Spreading Room, MCR air filters, I I and operator living area. l d. Diesel Generator Building - Corrider Area. e. Turbine Building - Control Building Wall. i APPLICABILITY: Whenever equipment protected by the spray / sprinkler system is o required.to be OPERABLE. ACTION: j. With one or more of the above required rpray and/or sprinkler systems a. inoperable, within one hour establish a continuous fire watch with l backup fire suppression equipment for those areas in which redundant i systems or components could be damaged; for other areas establish an hourly fire watch patrol. Restore the system to OPERABLE status within 14 days or, in lieu of any other report required by Specifica-tion 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status. b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable. 2 SURVEILLANCE REQUIREMENTS l i 4.7.11.2 Each of the above required spray and/or sprinkler systems shall be demonstrated OPERABLE: At least once per 31 days by verifying that each valve (manual, a. power operated or automatic) in the flow path is in its correct position. b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel. SEQUOYAH - UNIT 1 3/4 7-33 Amendment No.12 i = ~ . ~..... - - - -.. -.,. - - - -=

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c. At least once per 18 months: 1. By performing a system functional test which includes simulated automatic actuation of the system, and: a) Verifying that the automatic valves in the flow path actuate to their correct positions on a cross zone or single zone detection test signal as designed, and b) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel. 2. By visual inspection of the dry pipe, spray and sprinkler headers to verify their integrity, and 3. By visual inspection of each nozzle's spray area to verify the spray pattern is not obstructed. SEQUOYAH - UNIT 1 3/4 7-34 Amendment No. 12

PLANT SYSTEMS i _C_O,,SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.11.3 The following low pressure CO systems shall be OPERABLE. 2 t a. Cable Spreading Room. I b.' Computer Room. c. Auxiliary Instrument Room. d. Diesel Generator Rooms. e. Fuel Oil Pump Rooms. APPLICABILITY: Whenever equipment protected by the CO systems is required to be OPERABLE. 2 ACTION: a. With one or more of the above required CO systems inoperable, 2 within one hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol. Restore the system to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.7.11.3.1 Each of the above required CO systems shall be demonstrated 2 OPERABLE at least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path is in its correct position. t SEQUOYAH - UNIT 1 3/4 7-35~ Amendment No.12

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.7.11.3.2 Each of the above required low pressure CO systems shall be 2 demonstrated OPERABLE: a. At least once per 7 days by verifying the CO st rage tank level to 2 be greater than 50% and pressure to be greater than 270 psig, and b. At least once per 18 months by verifying: 1. The system valves and associated ventilation dampers and fire door release mechanisms actuate manually and automatically, upon receipt of a simulated actuation signal, and 2. Flow from each nozzle during a " Puff Test." SEQUOYAH - UNIT 1 3/4 7-36 Amendment No. 12

f PLANT SYSTEMS FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION 3.7.11.4 The fire hose stations shown in Table 3.7-10 shall be OPERABLE. _ APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE. i ACTION: a. With one or more of the fire hose stations shown in Table 3.7-10 inoperable, route an additional equivalent capacity fire hose to the unprotected area (s) from an OPERABLE hose station within I hour if the inoperable fire hose is the primary means of fire suppression; otherwise route the additional hose within 24 hours. Restore the fire hose station to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the station to OPERABLE status. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.7.11.4 Each of the fire hose stations shown in Table 3.7-10 shall be demonstrated OPERABLE: a. At least once per 31 days by visual inspection of the stations accessible during plant operations to assure all required equipment is at the station. b. At least once per 18 months by: 1. Visual inspection of the stations not accessible during plant operations to assure all recoired equipment is at the station, 2. Removing the hose for inspection and re-racking, and 3. Inspecting all gaskets and replacing any degraded gaskets in the couplings. c. At least once per 3 years by: 1. Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage. 2. Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pressure, whichever is greater. SEQUOYAH - UNIT 1 3/4 7-37 Amendment No.12 j-b y-a s 4 e-vM-3'#-%Wf W D'S 3

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s PLANT SYSTEMS s ) TABLE 3.7-10 4 F_ IRE HOSE STATIONS LOCATION ELEVATION HOSE RACK # a. Reactor Building - Annulus Area Platform 778.5 1-26-1196 Platform 778.5 1-26-1197 Platform 778.5 1-26-1198 Platform 778.5 1-26-1199 Platform 759.5 1-26-1200 Platform 759.5 1-26-1201 Platform 759.5 1-26-1202 Platform 759.5 1-26-1203 Platform 740.5 1-26-1204 Platform 740.5 1-26-1205 Platform 740.5 1-26-1206 Platform 740.5 1-26-1207 Platform 721.5 1-26-1208 Platform 721.5 1-26-1209 Platform 721.5 1-26-1210 ' Platform 721.5 1-26-1211 Platform 701.5 1-26-1212 Platform 701.5 1-26-1213 Platform 701.5 1-26-1214 Platform 701.5 1-26-1215 Platform 679.78 1-26-1216 Platform 679.78 1-26-1217 Platform 679.78 1-26-1218 Platform 679.78 1-26-1219 b. Reactor Building - RCP & Lower Containment Air Filters Area Reactor Building 679.78 1-26-1220 Reactor Building 679.78 1-26-1221. Reactor Building 679.78 1-26-1222 Reactor Building 679.78 1-26-1223 Reactor Building 679.78 1-26-1224 Reactor Building 679.78 1-26-1225 c. Control Building Control Building 732 0-26-1186 Control Building 732 0-26-1191 Control Building 706 0-26-1187 Control Building 706 0-26-1192 i g SEQUOYAH - UNIT 1 3/4 7-38 Amendment No. 12

1 TABLE 3.7-10 (Continued) FIRE HOSE STATIONS LOCATION ELEVATION HOSE RACK # Control Building 685 0-26-1188 Control Building 685 0-26-1193 Control Building 669 0-26-1189 Control Building 669 0-26-1194 d. Diesel Generator Building Corridor 722 0-26-1077 Corridor 740.5 0-26-1080 Air Exhaust Rm. 0-26-1082 Additional Equipment Building - Unit 1 e. South Wall 740.5 1-26-687 South Wall 706 1-26-686 f. Auxiliary Building 759 1-26-669 749 2-26-664 749 1-26-664 734 2-26-670 734 0-26-684 734 1-26-670 734 0-26-682 734 1-26-671 Siamese Outlet 734 1-26-672 734 1-26-665 714 0-26-660 714 1-26-666 714 0-26-677 706 0-26-658 690 0-26-690 690 0-26-661 690 1-26-674 Siamese Outlet 690 1-26-675 690 1-26-667 669 1-26-668 669 0-26-662 669 0-26-680 653 0-26-663 653 0-26-691 SEQUOYAH - UNIT 1 3/4 7-39 Amendment No. 12

TABLE 3.7-10 (Continued) FIRE HOSE STATIONS LOCATION ELEVATION HOSE RACK # g. CCW Intake Pumping Station 690 0-26-866 690 0-26-867 690 0-26-868 690 0-26-869 690 0-26-870 h. ERCW Pumping Station 688 0-26-927 688 0-26-926 688 0-26-930 704 0-26-931 704 0-26-925 704 0-26-928 720 0-26-929 720 0-26-924 720 0-26-932 SEQUOYAH - UNIT 1 3/4 7-40 Amendment No. 12

PLANT SYSTEMS 3/4.7.12 FIRE BARRIER PENETRATIONS LIMITING CONDITION FOR OPERATION 3.7.12 All fire barrier penetrations (including cable penetration barriers, firedoors and fire dampers) in fire zone boundaries protecting safety related areas shall be functional. APPLICABILITY: At all times. ACTION: With one or more of the above required fire barrier penetrations a. non-functional, within one hour either, establish a continuous fire watch on at least one side of the affected penetration, or verify the OPERABILITY of fire detectors on at least one side of the non-functional fire barrier and establish a hourly fire watch patrol. Restore the non-functional fire barrier penetration (s) to functional status within 7 days or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the non-functional pene-tration and plans and schedule for restoring the fire barrier penetration (s) to functional status. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.7.12 Each of the above required fire barrier penetrations shall be verified to be functional: At least once per 18 months by a visual inspection a. b. Prior to returning a fire barrier penetration to functional status following repairs or maintenance by performance of a visual inspec-tion of the affected fire barrier penetration (s). SEQUOYAH - UNIT 1 3/4 7-41 Amendment No. 12 I L

_~ i i 3 i i;;. m FC":Ut 5t5if.M5-i ~ l 6 EtN N !!ME Rf 091RREN!S (Cogti pd)_ _ _ i f 4. Voci fying the diesel it arts from ambient condition and - ( accelerates to at least 900 rpm in less than or equal to 10 seconds. The generator voltage and frequency shall oc ( 6900 t 690 volts and 60 1.2 Ilz within 10 seconds af ter u;n start signal. The diesel generator shall be started for th.is [ ] tist by using one of the folicwing signals with start op on cach q signal verified at least once per 124 days: i j a) Manual. 1 i j b) Sinuidted loss of offsite power by itself.

t-I c)

Simulated loss of offsite pov.er in conjunction with actuation test signal. I { d) An ESF actation test sgnal by itself. l S. Verifying the generator is synchronized, loaded to groater t%n } qual to 4000 kw in in:s t' in or aqual to 60 wcendi, c.d c, rdtes for greater than or equal to 60 minutes, and i i i f,. W rifying the diesel-generator is aligned to previde standby w ar to the associated shutdown boards. i t j b. At !ast once per 31 days and af ter each operation of the diesel ,i th period or opacatico was greater t%n or qual to I hou: 4 ]

q Ncking for and removing accaulated wter frem the engine-e ned fuel tenks.

7 J i c. At 'least once. per 92 days and frra epw fuel oil prior ta eddition to t he 7-day tanks by verifying tb it a segle ehtained in accor_danen 4 Ath Antlul0-19/5 has a water and wdicent datet or L n than or qual to. CS volume percent. and a' kinematic vinccr.i ty 0 kn*F of i

rMer than or Squal to~ 1.8 but less than.or squM L: o E 3;cohti-j

_,' u in trmt6d in accordance with ASTM-0975-//, and an ppucity kmi.of 7 ass than 2. mg. of insolubles per 100'ml. when testhd in t j

cuiia sb uith ASTM 402274-70.

{ At _ b nt a.se. per 18 months during shutdown byi d. t 1. %jecting the diesel to an inspection in accordance with I peujut es prepared in cenjunction with its nauracturer's i ,w inditions for this class of-standby '.c ;;ca; t j 7. Wrify!pg the generator capability to reject a ; lead of gPeatcr j th ur or tqual to-600 kw While ruintaining ' volt;cy at rM001690 n i volts Jnd fr gency at 6011;2 lu. 1 4 I 3. Wri fying the generator q;iability La reject a lo4d 00 4p00 tt Without' tripping, lhe generster voltage shall L nut -read 4 .)C E volts dori d and following the load rejection. .-S Q yffMl - Ldif j-3M 3-3 Ah "8st No. 12 I

l ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) { 4. Simulating a loss of offsite power by itself, and: a) Verifying de-energization of the shutdown boards and load shedding from the shutdown boards, b) Verifying the diesel starts on the auto-start signal, energizes the shutdown boards with permanently connected loads within 10 seconds, energizes the auto-connected shutdown loads through the load sequencers and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady state voltage and frequency of the shutdown boards shall be maintained at 6900 690 volts and 60 1.2 Hz during this test. 5. Verifying that on a ESF actuation test signal (without loss of offsite power) the diesel generator starts on the auto-start signal ar.d operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 6900 690 volts and 60 1 1.2 Hz within 10 seconds after the auto-start signal; the steady state generator voltage and fre-quency shall be maintained within these limits during this test. 6. Verifying that on a simulated loss of the diesel generator (with offsite power not available), the loads are shed from the shutdown boards and that subsequent loading of the diesel generator is in accordance with design requirements. 3 7. Simulating a loss of offsite power in conjunction with an ESF actuation test signal, and a) Verifying de energization of the shutdown boards and load shedding from the shutdown boards. b) Verifying the diesel starts from ambient condition on the auto start signal, energizes the shutdown boards with permanently connected loads within 10 seconds, energizes the auto-connected emergency (accident) loads through the load sequencers and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. Af ter energization, the steady state voltage and frequency of the emergency busses shall be maintained at 6900 690 volts and 60 1.2 Hz during this test. c) Verifying that all automatic diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on the shutdown board and/or safety injection actuation signal. 8. Verifying the diesel generator operates for at least 24 hours. During the first 2 hours of this test, the diesel generator shall be loaded to greater than or equal to 4400 kw and during the remaining 22 hours of this test, the diesel generator shall be loaded to greater than or equal to 4000 kw. SEQUOYAH - UNIT 1 3/4 8-4 Amendment No. 12

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) Within 5 minutes after completing this 24 hour test, perform ~ Specification 4.8.1.1.2.c.4. The generator voltage and fre-quency shall be 6900 1 690 volts and 60 i 1.2 Hz within 10 seconds after the start signal; the steady state generator voltage and frequency shall be maintained within these limits during this test. 9. Verifying that the auto-connected loads to each diesel generator do not exceed the 2000 hour rating of 4000 kw. 10. Verifying the diesel generator's capability to: a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power. b) Transfer its loads to the offsite power source, and c) Be restored to its shutdown status. 11. Verifying that with the diesel generator operating in a test mode (connected to its bus), a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) automatically energizing the emergency loads with offsite power. 12. Verifying that the automatic load sequence timers are OPERABLE with the setpoint for each sequence timer within + 5 percent of its design setpoint. 13. Verifying that the following diesel generator lockout features prevent diesel generator starting only when required: a) Engine overspeed b) 86 GA lockout relay At least once per 10 years or af ter any modifications which could e. affect diesel generator interdependence by starting the diesel generators simultaneously, during shutdown, and verifying that the diesel generators accelerate to at least 900 rpm in less than or equal to 10 seconds, f. At least once per 10 years

by:

1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypoclorite solution, and 2. Performing a pressure test of those portions of the diesel fuel oil system design to Section III, subsection ND of the ASME Code at a test pressure equal to 110 percent of the system design pressure.

These requirements are waived for the initial surveillance.

SEQUOYAH - UNIT 1 3/4 8-5 Amendment No. 12 ~

i ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i 4.8.1.1.3 The 125 volt D.C. distribution panel, 125-volt D.C. battery bank and associated charger for each diesel generator shall be demonstrated OPERABLE: f a. At least once per 7 days by verifying: J \\ 1. That the parameters in Table 4.8-la meet the Category A limits. l 2. That the total battery terminal voltage is greater than or equal to 129-volts on float charge. b. At least once per 92 days by: 1. Verifying that the parameters in Table 4.8-la meet the i Category B limits, i 2. Verifying there is no visible corrosion at either terminals or connectors, or the cell to terminal connection resistance of these items is less than 150 x 10 6 ohms, and 3. Verifying that the average electrolyte temperature of 6 connected cells is above 60 F. I i At least once per 18 months by verifying that: c. 1. The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration. 2. The battery to battery and terminal connections are clean, tight and coated with anti-corrosion material. i 3. The resistance of each cell to terminal connection is less i than or equal to 150 x 10 6 ohms. 4.8.1.1.4 Repcets - All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.1. Reports of I diesel generator failures shall include the information recommended in Regula-tory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the number of failures in the last 100 valid tests (on a per nuclear unit basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. ~ j l' l L SEQUOYAll - UNIT 1 3/4 8-6 Amendment No. 12 L f e ~ ..,,,,n,..

I TABLE 4.8.la DIESEL GENERATOR BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A(1) CATEGORY B(2) Parameter Limits for each Limits for each Allowable (3) designated pilot connected cell value for each cell connected cell Electrolyte > Minimum level > Minimum level Above top of Level indication mark, indication mark, plates, and and < %" above and < %" above not overflowing maxi 3um level maxiEum level indication mark indication mark r Float Voltage > 2.13 volts > 2.13 volts (c) > 2.07 volts Not more than .020 below the average of all [ > 1.190 connected cells 1 1.195(b) fP a) r Average of all Average of all connected cells >1.190{g) cells connect > 1.200 (a) Corrected for electrolyte temperature and level. (b) Or battery charging current is less than 2 amps. (c) Corrected for average electrolyte temperature. (1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and i provided all parameter (s) are restored to within limits within the next 6 days. (2) For any Category B parameter (s) outside the limit (s) shown, the battery l may be considered OPERABLE provided that they are within their allowable values and provided the parameter (s) are restored to within limits within 7 days. (3) Any Category B parameter not within its allowable value indicates an inoperable battery. b SEQUOYAH - UNIT'1 3/4 8-7a Amendment No. 12

l ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 The following A.C. electrical boards shall be OPERABLE and energized l with tie breakers open between redundant boards: 6900 Volt Shutdown Board 1A-A 6900 Volt Shutdown Board 18-8 6900 Volt Shutdown Board 2A-A 6900 Volt Shutdown Board 28-8 480 Volt Shutdown Board 1Al-A l 480 Volt Shutdown Board 1A2-A 480 Volt Shutdown Board IB1-B t 480 Volt Shutdown Board 182-8 480 Volt Shutdown Board 2Al-A l 480 Volt Shutdown Board 2A2-A 480 Volt Shutdown Board 281-B 480 Volt Shutdown Board 282-B 120 Volt A.C. Vital Instrument Power Board Channels 1-I and 2-I energized from inverters 1-I and 2-I connected to D.C. Channel I*. 120 Volt A.C. Vital Instrument Power Board Channels 1-II and 2-II energized from inverters 1-II and 2-II connected to D.C. Channel II*. 120 Volt A.C. Vital Instrument Power Board Channels 1-III and 2-III energized from inverters 1-III and 2-III connected to D.C. Channel III*. 120 Volt A.C. Vital Instrument Power Board Channels 1-IV and 2-IV energized from inverters 1-IV and 2-IV connected to D.C. Channel IV*. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With less than the above complement of A.C. boards OPERABLE and energized, a. restore the inoperable boards to OPERABLE status within 8 hours or be in [ at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within i the following 30 hours. i b. With one inverter inoperable, energize the associated Vital Instrument Power Board within 8 hours; restore the inoperable inverter to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.8.2.1 The specified A.C. boards and inverters shall be determined OPERABLE and energized with tie breakers open between redundant boards at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.

Two inverters may be disconnected from their D.C. source for up to 24 hours for the i

purpose of performing an equalizing charge on their associated battery bank provide (1) the vital instrument power board is OPERABLE and energized, and (2) the vital l instrument power boards associated with the other battery banks are OPERABLE and energized from their respective inverters connected to their respective D.C. source. SEQUOYAH - UNIT 1 3/4 8-9 Amendment No. 12 l

ELECTRICAL POWER SYSTEMS A.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, the following A.C. electrical boards and inverters shall be OPERABLE and energized: 2-6900 volt shutdown boards, either lA-A and 2A-A or 1B-B and 2B-B, 4-480 volt shutdown boards associated with the required OPERABLE 6900 volt shutdown boards, 2-120 volt A.C. vital instrument power boards either Channels I and III or Channels II and IV energized from their respective inverters connected to their respective D.C. battery banks, and 480 volt shutdown boards. APPLICABILITY: MODES 5 and 6. ACTION: With less than the above c 'plement of A.C. boards and inverters OPERABLE and energized, establish CONTA4.iENT INTEGRITY within 8 hours. SURVEILLANCE REQUIREMENTS 4.8.2.2 The specified A.C. boards and inverters shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and indicated voltage on the bus. SEQUOYAH - UNIT 1 3/4 8-10 Amendment No. 12

l ELECTRICAL F0WER SYSTEMS l / D.C. DISTRIBUTION - OPERATING \\ LIMITING CONDITION FOR OPERATION 3.S The following D.C. vital battery channels shall be energized and OPEh,.ot e: CHANNEL I Consisting of 125 -volt D.C. board No. I,125 - volt D.C. battery bank No. I and a full capacity charger. CHANNEL II Consisting of 125 - volt D.C. board No. II, 125 - volt D.C. battery bank No. II, and a full capacity charger. CHANNEL III Consisting of 125 - volt D.C. board No. III, 125 - volt D.C. battery bank No. III, and a full capacity charger. CHANNEL IV Consisting of 125 - volt D.C. board No. IV,125 - volt D. C. battery bank No. IV, and a full capacity charger. APPLICABILITY: M3 DES 1, 2, 3 and 4. ACTION: With one 125-volt D.C. board inoperable, restore the inoperable a. board to OPERABLE status within 2 hours or be in at least HOT STANDBY ( within the next 6 hours and in COLD SHUTbOWN within the following 30 hours. b. With one 125-volt D.C. battery bank and/or its charger inoperable, restore the inoperable battery bank and/or charger to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. x SEQUOYAH - UNIT 1 3/4 8-11

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS l 4.8.2.3.1-Each D.C. bus train shall be determined OPERABLE and energized with tie breakers open between redundant busses at least once per 7 days by verifying correct breaker alignment, indicated power availability from the charger and battery, and voltage on the bus of greater than or equal to 125 volts. l 4.8.2.3.2 Each 125 volt battery bank and charger shall be demonstrated OPERABLE: a a. At least once per 7 days by: l 1. Verifying that the parameters in Table 4.8-2 meet the [ Category A limits, and 2. Verifying total battery terminal voltage is greater than or equal to 129-volts on float charge. b. At leart once per 92 days and within 7 days after a battery discharge (battery terminal voltage below 110-volts), or battery overcharge [ (battery terminal voltage above 150-volts), by: l 1. Verifying that the parameters in Table 4.8-2 meet the Category B

limits, 2.

Verifying there is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10 6 ohms, and 3. Verifying that the average electrolyte temperature of 6 connected i cells is above 60 F. At least once per 18 moaths by verifying that: c. - i 1. The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration, 2. The cell-to-cell and terminal connections are clean, tight and coated with anti-corrosion material, 3. ' Tne resistance of each cell-to-terminal connection is less than or equal to 150 x 10 G ohms, and 4. The battery charger will supply at least 150 amperes at 125 volts for at least 4 hours. t SEQUOYAH - UNIT 1 3/4 8-12 Amendment No. 12 q w, . wm. ,=#-ww++r-ev--- mw - re--

j - ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) d. At least once per.18 months by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for 2 hours when the battery is subjected to a battery service test. I At least once per 60 months by verifying that the battery capacity e. is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Once per 60 month interval, this i performance discharge test may be performed in lieu of the battery service test. f. Annual performance discharge tests of battery capacity shall be given to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating. a 4 l t i 4 4 J SEQUOYAH - UNIT,1 3/4 8-13 Amendment No. 12

t TABLE 4.8.2 BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A(1) CATEGORY B(2) Parameter Limits for each Limits for ach Allowable (3) designated pilot connected cell value for each cell connected cell Electrolyte > Minimum level > Minimum level Above top of Level indication mark, indication mark, plates, and and 1 b" above and 5 k" above not overflowing maximum level maximum level indication mark indication mark Float Voltage > 2.13 volts > 2.13 volts (c) > 2.07 volts Not more than .020 below the average of all > 1.195 connected cells Specifi > 1.200(b) Gravity [a) connected cells >1.195{g) cells connect > 1.205 (a) Corrected for electrolyte temperature and level. (b) Or battery charging current is less than 2 amps. (c) Corrected for average electrolyte temperature. (1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category 8 measurements are taken and found to be within their allowable values, and provided all parameter (s) are restored to within limits within the next 6 days. (2) For any Category B parameter (s) outside the limit (s) shown, the battery ~ may be considered OPERABLE provided that they are within their allowable values and provided the parameter (s) are restored to within limits within 7 days. (3) Any Category B parameter not within its allowable value indicates an inoperable battery. SEQUOYAH - UNIT 1 3/4 8-13a Amendment No. 12

ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION - SHUTDOWN -('- LIMITING CONDITION FOR OPERATION 3.8.2.4 As a minimum, the following D.C. electrical equipment and boards shall be energized and OPERABLE: 2-125-volt D.C. boards either I and III or II and IV, and 2-125-volt battery banks and chargers, one associated with each operable D.C. board APPLICABILITY: MODES 5 and 6. ACTION: With less than the above complement of D.C. equipment and board OPERABLE, establish CONTAINMENT INTEGRITY within 8 hours. SURVEILLANCE REQUIREMENTS 4.8.2.4.1 The above required 125-volt D.C. vital battery boards shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and indicated power availability with an overall battery voltage of greater than or equal to 125 volts. 4.8.2.4.2 The above required 125-volt D.C. vital battery banks and chargers shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2. ( SEQUDYAH - UNIT 1 3/4 8-14

E TABLE 3.8-2 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION AND/0R BYPASS DEVICES Valve No. Function Bypass Device 1-FCV-62-63 Isolation for Seal Water Filter No 1-FCV-62-138 Safe Shutdown Redundancy (CVCS) No 1-FCV-62-98 ECCS Operation No 1-FCV-62-99 ECCS Operation No 1-FCV-62-90 ECCS Operation No 1-FCV-62-91 ECCS Operation No 1-FCV-62-61 Cont. Isolation No 1-LCV-62-132 ECCS Operation No 1-LCV-62-133 ECCS Operation No 1-LCV-62-135 ECCS Operation No 1-LCV-62-136 ECCS Operation No 1-FCV-74-1 Open for Normal Plant Cooldown No 1-FCV-74-2 Open for Normal Plant Cooldown No 1-FCV-74-3 ECCS Operation No 1-FCV-74-21 ECCS Operation No 1-FCV-74-12 RHR Pump, Mini-flow Protects Pump No 1-FCV-74-24 RHR Pump, Mini-flow Protects Pump No 1-FCV-74-33 ECCS Operation No 4 1-FCV-74-35 ECCS Operation No 1-FCV-74-7 ECCS Operation No 1-FCV-74-6 ECCS Operation No 1-FCV-63-156 ECCS Flow Path No 1-FCV-63-157 ECCS Flow Path No 1-FCV-63-39 BIT Injection No 1-FCV-63-40 BIT Injection No 1-FCV-63-25 BIT Injection No 1-FCV-63-26 BIT Injection No 1-FCV-63-118 RCS Pressure Boundary No 1-FCV-63-98 RCS Pressure Boundary No 1-FCV-63-80 RCS Pressure Boundary No 1-FCV-63-67 RCS Pressure Boundary No 1-FCV-63-1 ECCS Operation No 1-FCV-63-72 ECCS Flow Path from Cont. Sump No 1-FCV-63-73 ECCS Flow Path from Cont. Sump No 1-FCV-63-8 ECCS Flow Path No 1-FCV-63-11 ECCS Flow Path No 1-FCV-63-93 ECCS Cooldown Flow Path No 1-FCV-63-94 ECCS Cooldown Flow Path No 1-FCV-63-172 ECCS Flow Path No 1-FCV-63-5 ECCS Flow Path No 1-FCV-63-47 Train Isolation No 1-FCV-63-48 Train Isolation No 1-FCV-63-4 SI Pump Mini-flow No 1-FCV-63-175 SI Pump Mini-flow No SEQUOYAH - UNIT 1 3/4 8-35 Amendment No. 12 - ~

t i TABLE 3.8-2 (Continued) MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION AND/0R BYPASS DEVICES Valve No. Function Bypass Device ~ ~ l 1-FCV-63-3 SI Pump Mini-flow No 1-FCV-63-152 ECCS Recirc No i 1-FCV-63-153 ECCS Recirc No 1 1-FCV-63-22 ECCS Recirc No l 1-FCV-3-33 Quick Closing Isolation No 1-FCV-3-47 Quick Closing Isolation No 1-FCV-3-87 Quick Closing Isolation No l 1-FCV-3-100 Quick Closing Isolation No 1-FCV-1-15 Stm Supply to Aux FWP turbine No 1-FCV-1-16 Stm Supply to Aux FWP turbine No 1-FCV-3-179A ERCW Sys Supply to Pump No 4 1-FCV-3-1798 ERCW Sys Supply to Pump No i 1-FCV-3-136A ERCW Sys Supply to Pump No l-1-FCV-3-136B ERCW Sys Supply to Pump No 1-FCV-3-116A ERCW Sys Supply to Pump No 1-FCV-3-116B ERCW Sys Supply to Pump No 1-FCV-3-126A ERCW Sys Supply to Pump No i 1-FCV-3-126B ERCW Sys Supply to Pump No 1 1-FCV-70-133 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-139 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-4 Isolation for Non-Essential Loads No 1-FCV-70-143 Isolation for Excess Letdown Ht Xchngr No 1-FCV-70-92 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-90 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-87 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-89 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-140 Isolation for RCP Oil Coolers & Therm B No 1-FCV-70-134 Isolation for RCP Oil Coolers & Therm B No 1-FCV-67-67*. DG Ht Ex No 2-FCV-67-65* DG Ht Ex No 1-FCV-67-66* DG Ht Ex No 2-FCV-67-68* DG Ht Ex No 1-FCV-67-123 CSS Ht Ex Supply No 1-FCV-67-125 CSS Ht Ex Supply No 1-FCV-67-124 CSS Ht Ex Discharge No i 1-FCV-67-126 CSS Ht Ex Discharge No i 0-FCV-67-151* CCW Ht Ex Throttling No 0-FCV-67-152* CCW Ht Ex Throttling No 2-FCV 146 CCW Ht Ex Throttling No 2-FCV-67-223. Isolation of 1B/2A HDR's No i 1-FCV-67-83 Cont. Isol. Lower No 1-FCV-67-88 Cont. Isol. Lower -No 1-FCV-67-87 Cont. Isol. Lower No 1-FCV-67-424* CCW Ht Ex Isolation No 1-FCV-67-478* Isolation of IB ERCW HDR No 4

Common to Units 1 & 2 SEQUOYAH - UNIT 1 3/4 8-36 Amendment No.

12 a..-,,_._

k TABLE 3.8-2 (Continued) MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION AND/OR BYPASS DEVICES Valve No. Function Bypass Device 1-FCV-67-95 Cont. Isol. Lower No 1-FCV-67-96 Cont. Isol. Lower No 1-FCV-67-91 Cont. Isol. Lower No 1-FCV-67-103 Cont. Isol. Lower No 1-FCV-67-104 Cont. Isol. Lower No 1-FCV-67-99 Cont. Isol. Lower No 1-FCV-67-lll Cont. Isol. Lower No i 1-FCV-67-ll2 Cont. Isol. Lower No 1-FCV-67-107 Cont. Isol. Lower No 1-FCV-67-130 Cont. Isol. Upper No 1-FCV-67-131 Cont. Isol. Upper No 1-FCV-67-295 Cont. Isol. Upper No 1-FCV-67-134 Cont. Isol. Upper No 1-FCV-67-296 Cont. Isol. Upper No 1-FCV-67-133 Cont. Isol. Upper No 1-FCV-67-139 Cont. Isol. Upper No 1-FCV-67-297 Cont. Isol. Upper No 1-FCV-67-138 Cont. Isol. Upper No d 1-FCV-67-142 Cont. Isol. Upper No 1-FCV-67-298 Cont. Isol. Upper No 1-FCV-67-141 Cont. Isol. Upper No 1-FCV-72-21 Cont. Spray Pump Suction No 1-FCV-72-22 Cont. Spray Pump Suction No 1-FCV-72-44 Cont. Spray Pump Suction No 1-FCV-72-45 Cont.- Spray Pump Suction No 1-FCV-72-2 Cont. Spray Isol. No 1-FCV-72-39 Cont. Spray Isol. No 1-FCV-72-40 RHR Cont. Spray Isol. No 1-FCV-72-41 RHR Cont. Spray Isol. No t i 4 I SEQUOYAH - UNIT 1-3/4 8-37 Amendment No. 12

ELECTRICAL POWER SYSlEMS ISOLATION DEVICES 4 LIMITING CONDITION FOR OPERATION -a 3;8.3.3 All circuit b. eakers actuated by fault currents that are used as n W isolation devices protecting IE busses from non qualified loads shall be OPERABLE. APPLICABIlllY: MODES 1, 2, 3 and 4. ~ ACTION: i I With one or more of the above required circuit breakers inoperable either: Restore the inoperable circuit breaker (s) to OPERABLE status within a. 8 hours, or b. Trip the inoperable circuit breaker (s), rack-out the circuit breake'r(s) within 8 hours and verify the circuit breaker (s) to be racked out at least once per 7 days thereaf ter; the provisions of Specification 3.0.4 are not applicable to racked-out circuit breakers, or Be in at least HOT STANDBY within the next 6 hours and in COLD c. ' SHUTDOWN within the following 30 hours. i I SURVEILLANCE REQUIREMENTS 4.8.3.3 Each of the above required circuit breakers shall be demonstrated i OPERABLE: } J j a. At least once per 18 months by selecting and functionally testing a representative sample of at least 10% of each type of circuit breaker. j Circuit breakers selected for functional testing shall be selected L on a rotating basis. i The functional test shall consist of. injecting a current input at the specified setpoint to.each selected circuit i breaker or relay and verifying that each circuit breaker functions as designed. For each device found inoperable during these functional I tests, an additional representative sample of at least 10% of each i over current protection device of the inoperable type shall also be functionally tested until no more failures are found or all devices of that type have been functionally tested. b. At least once per 60 months by subjecting each circuit breaker to an-1 inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations. I 1 SEQUOYAH - UNIT 1 3/4 8-38 =0 n ..r ,.,.4

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head closure bolts less than fully tensioned or ~ with the head removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met: a. Either a K f 0.95 or less, which includes a 1% delta k/k conser-eff vative allowance for uncertainties, or b. A boron concentration of greater than or equal to 2000 ppm, which includes a 50 ppm conservative allowance for uncertainties. APPLICABILITY: MODE 6* ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or its equivalent until K is reduced to less than or equal to 0.95 or the boron eff concentration is restored to greater than or equal to 2000 ppm, whichever is the more restrictive. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS

4. 9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

Removing or unbolting the reactor vessel head, and a. b. Withdrawal of any full lerafd control rod in excess of 3 feet from its fully inserted posit'or eithin the reactor pressure vessel. 4.9.1.2 The boron concentratFO 'ti reactor coolant system and the refueling canal shall be determined by ch ; t snalysis at least once per 72 hours.

4. 9.1. 3 One of the following zalve combinations shall be verified closed under administrative control at least once per 72 hours:

Combination A Combination B Combination C Combination D a. 1-81-536 a. 1-81-536 a. 1-81-536 a. 1-81-536 b. 1-62-922 b. 1-62-922 b. 1-62-907 b. 1-62-907 c. 1-62-916 c. 1-62-916 c. 1-62-914 c. 1-62-914 d. 1-62-933 d. 1-62-940 d. 1-62-921 d. 1-62-921 e. 1-62-696 c. 1-62-933 e. 1-62-940 f. 1-62-929 f. 1-62-929 g. 1-62-932 g. 1-62-932 i

h..1-FCV-62-128 h.

1-62-696 i. 1-FCV-62-128

The reactor shall be maintained in MODE 6 whenever fuel is in the reactor i

vessel with the vessel head closure bolts less than fully tensioned or with L the head removed. SEQUOYAH -' UNIT 1 3/4 9-1 Amendment No.12 l l

REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be OPERABLE and uperating, each with continuous visual indication in the control room and one with audible indication in the containment and control room. APPLICABILITY: MODE 6. ACTION: With one of the above required monitors inoperable or not operating, a. immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. b. With both of the above required monitors inoperable or not operating, determine the boron concentration of the reactor coolant system at least once per 12 hours. The provisions of Specification 3.0.3 are not applicable. c. SURVEILLANCE REQUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of: A CHANNEL CHECK at least once per 12 hours, a. b. A CHANNEL FUNCTIONAL TEST at least once per 7 days, and A CHANNEL FUNCTIONAL TEST within 8 hours prior to the initial start c. of CORE ALTERATIONS. SEQUOYAH - UNIT 1 3/4 9-2 Amendment No. 12 L. -_...m....

REFUELING OPERATIONS [ 3/4.9.3 DEC'AY TIME LIMITING CONDITION FOR OPERATION J 3.9.3 The reactor shall be subcritical for at least 100 hours. A_PPLICABILITY: During movement of irradiated fuel in the reactor pressure vessel. e ACTION: With the reactor subtritical for less than 100 hours, suspend all operations involving movement of' irradiated fuel in the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable. [ SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least 100 hours by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel. e SEQUOYAH - UNIT 1 3/4 9-3

REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS i-LIMITING CONDITION FOR OPERATION l 3.9.4 The containment building penetrations shall be in the following status: The equipment door closed and held in place by a minimum of four a.

bolts, b.

A minimum of one door in each airlock is closed, and c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either: 1. Closed by an isolation valve, blind flange, or manual valve, or 2. Be capable of being closed by an OPERABLE automatic Containment Ventilation isolation valve. APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: With the requirements of the above specification not satisfied, immediately ] suspend all operations involving CORE ALTERATIONS or movement of irradiated i fuel in the containment building. The provisions of Specification 3.0.3 are not applicable. ] SURVEILLANCE REQUIREMENTS 4.9.4 Each of the above required containment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERABLE automatic Containment Ventilation isolation valve within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by: Verifying the penetrations are in their closed / isolated a. condition, or b. . Testing the Containment Ventilation isolation valves per the applicable portions of Specification 4.6.3.2. I J d i

SEQUOYAH - UNIT 1 3/4 9-4 Amendment No. 12

REFUELING OPERATIONS 3/4.9.5 COMMUNICATIONS / i LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station. APPLICABILITY: During CORE ALTERATIONS. ACTION: When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the start of and at least once per 12 hours during CORE ALTERATIONS. SEQUOYAH - UNIT 1 3/4 9-5

REFUELING OPERATIONS 3/4.9.6 MANIPULATOR CRANE i LIMITING CONDITION FOR OPERATION a ~3.9.6 The manipulator crane and auxiliary hoist shall be used for movement of . drive rods or fuel assemblies and shall be OPERABLE with: The manipulator crane used for movement of fuel assemblies having: 7 a. 1. A minimum capacity of 2750 pounds, and 2. An overload cut off limit less than or equal to 2700 pounds. b. The auxiliary hoist used for latching and unlatching drive rods having: 1 1. A minimum capacity of 610 pounds, and 2. A load indicator which shall be used to prevent lifting loads in excess of 600 pounds. APPLICABILITY: During movement of drive rods or fuel assemblies within the reactor pressure vessel. ACTION: I With the requirements for crane and/or hoist OPERABILITY not satisfied, suspend use of any inoperable mainipulator crane and/or auxiliary hoist from operations involving the movement of drive rods and fuel assemblies within the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS l 4.9.6.1 Each manipulator crane used for movement of fuel assemblies within i the reactor pressure vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least 2750 pounds and demonstrating an automatic load cut off when the crane load exceeds 2700 pounds. 4.9.6.2 Each auxiliary hoist and associated load indicator used for movement of drive rods within the reactor pressure vessel shall be demonstrated OPERABLE within'100 hours prior to the start of such operations by performing a load test of at least 610 pounds. .SEQUOYAH - UNIT 1 3/4 9-6 Amendment No.12

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent Residual Heat Removal (RHR) loops shall be OPERABLE.* APPLICABILITY: MODE 6 when the water level above the top of the reactor pressure vessel flange is less than 23 feet.** ACTION: With less than the required RHR loops OPERABLE, immediately initiate a. corrective action to return the required RHR loops to OPERABLE status as soon as possible. b. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.8.2 The required Residual Heat Removal loops shall be determined OPERABLE per Specification 4.0.5.

The normal or emergency power source may be inoperable for each RHR loop.
- Prior to initial criticality only one independent RHR loop shall be required OPERABLE.

l SEQUOYAH - UNIT 1 3/4 9-8a Amendment No. 12

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITINC CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended f or measurement of control rod worth and shutdown margin provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s). APPLICABILITY: MODE 2. ACTION: With any full length control rod not f ully inserted and with less than a. the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. b. With all full length control rods inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length rod either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least 50% withdrawn position within 24 hours prior to reducing the SHUIDOWN MARGIN to less than the limits of Specification 3.1.1.1. I SEQUOYAH - UNIT 1 3/4 10-1 Amendment No. 12

. ~ t 4 SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS . LIMITING' CONDITION FOR OPERATION ~ i 3.10.2 The group height, insertion and power distribution limits of Specifica-tions 3.1.3.1, 3.1.3. 5, 3.1.3.6, 3.2.1 and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided: 'The THERMAL POWER is maintained less than or equal to 85% of RATED a. THERMAL POWER, and ~ b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below. a l APPLICABILITY: MODE 1 I ACTION: i { With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while t the requirements of Specification 3.1. 3.1., 3.1. 3. 5, 3.1. 3. 6, 3. 2.1 and 3. 2. 4 i are suspended, either: Reduce THERMAL POWER suf ficient to satisfy the ACTION requirements a. of Specifications 3.2.2 and 3.2.3, or } b. Be in HOT STANDBY within 6 hours. l SURVEILLANCE REQUIREMENTS ? .i 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. ) 4 4.10.2.2 Perform the surveillance required by the below listed Specifications { at-least once per 12 hours during PHYSICS TESTS: ~ a. Specification 4.2.2.2 and 4.2.2.3 5 .b. Specification 4.2.3.2. 2 SEQUOYAH -- UNIT 1 3/4 10-2 \\' .L

-. -..~_ ~ - - - iABLE 6.11-2 RADIDACTZVE GASEOUS BASTE MONITORING SAMPLING AND ANALYSIS PROGRAM -m Minimum Lower Limit of y Sampling Analysis Type of Detection (glD) g Gaseous Release Type Frequency Frequency Activity Analysis (pCi/ml) 'E P P 9 -4 s A. Waste Gas Storage Each Tank Each Tank Principal Gamma Emitters 1x10 s Tank Grab 5 Sample pi Di -4 B. Containment Purge Each Purge Each Purge Principal Gamma Emitters 9 lx10 Grab -6 Sample H-3 lx10 ~4 C. Noble Gases and M M Principal Gamma Emitters 9 lx10 i Tritium Grab 1.CondensgrVacuum Sample -6 11 - 3 lx10 Exhaust 2.Auxili Rebuilding Exhaust M

3. Service Building Exhaust y

4.Shieldgujlging e Exhaust D. Iodine and Parti-Continuous # W I-131 lx10 d -12 culates Sampler Charcoal -10

1. Auxiliary Building Sample I-133 lx10 Exhaust f

d Continuous W Principal Gamma Emitters 9 I lx 10

2. Shield Building Sampler Particulate (I-131, Others)

Exhaust Sample I Continuous M Gross Alpha lx10'II F Composite [ Sampler Particulate g Sample I Continuous Compbsite Sr-89, Sr-90 lx10'Il Er Sampler Particulate Sample w E. Noble Gases all Continuous # Noble Gas Noble Gases 1x10-6 Releases types as Monitor Monitor Gross Beta or Gamma listed in A, B, and C above 8 i

1 1ABLE 4.11-2 (Continued) TABLE NOTATION bc M a. The LLD is the smallest concentration of radioactive material in a sample 'that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radiochemical e separation): i 4.66 sb E V 2.22x106 Y exp (-Aat) Where: 4 LLD is the "a priori" lower limit of detection as defined above (as microcurie per unit mass or volume), is the standard deviation of the background counting rate or of sb tne counting rate of a blank sample as appropriate (as counts per minute), E is the counting efficiency (as counts per transformation), V is the sample size (in units of mass or volume), 2.22 x 106 is the number of transformations per minute per microcurie, I. Y is the fractional radiochemical yield (when applicable), A is the radioactive decay constant for the particular radionuclide, and-At is the elapsed time between midpoint of sample collect' ion and time of counting (for plant effluents, not environmental samples). The value of s used in the calculation of the LLD for a detection s counting rate or of the counting rate of the blank sam appropriata) rather than on an unverified theoretically predicted i Typical values of E, V, Y, and At shall be used in the var ane. calculation. i SEQUOYAH - UNIT 1 3/4 11-10 /. +

TABLE 4.11-2 (Continued) TABLE NOTATION b. Analyses shall also be performed following shutdown from 215% RATED THERMAL POWER, startup to 215% RATED THERMAL POWER or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within a one hour period. c. Tritium grab samples shall be taken at least once per 24 hours when the refueling canal is flooded. d. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing (or after removal from sampler). Sampling shall also be performea at least once per 24 hours for at least 2 days following each shutdown from 215% RATED THERMAL POWER, startup of 215% RATED THERMAL POWER or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in one hour and analyses shall be completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding LLD's may be increased by a factor of 10. Tritium grab samples shall be taken at least once per 7 days from e. the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool. f. The ratio of the sample flow rate to the sampled stream flow rate l shall be known for the tirae period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3. -g. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co 60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean tnat only these nuclides are to be detected and reported. Other peaks which are measureable and identifiable, together with the above nuclides, shall also be identified and report. h. During releases via this exhaust system. i. In MODES 1, 2, 3 and 4, the upper and lower compartments of the con-tainment shall be sampled prior to VENTING or PURGING. Prior to entering MODE 5, the upper and lower compartments of the containment shall be sampled. The incore instrument room purge sample shall be obtained at the shield building exhaust between 5 and 10 minutes following initiation of the incore instrument room purge. -1 SEQUOYAH - UNIT 1 3/4 11-11 Amendment No. 12 i 1

v I.ADICACTIVE ElFLUEhis DOSE - NOBLE GASES [ LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents from C the site (see Figure 5.1-1) shall be limited to the following*: During any calendar quarter: a. Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation and, b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation. APPLICABILITY: At all times. ACTION With the calculated air dose from radioactive noble gases in gaseous a. ef fluents exceeding any of the above limits, in lieu of any o+.her report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and . defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters, so that the cumulative dose is within 10 mrad for gamma ( radiation and 20 mrad for beta radiation. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once per 31 days. rPer reactor unit. s SEQUOYAH - UNIT 1 3/4 11-12 (

I t ~ i RADI0 ACTIVE EFFLUENTS I EXPLOSIVE GAS MIXTURE I i LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be j l limited to less than or equal to 2% by volume whenever the hydrogen f j concentration exceeds 4% by volume. >t i APPLICABILITY: At all times. I i i ACTION: I With the concentration of oxygen in a waste gas holdup tank greater i-a. than 2% by volume but less than or equal to 4% by volume, reduce the t oxygen concentration to the above limits within 48 hours. ) b. With the concentration of oxygen in a waste gas holdup tank greater than 4% by volume and the hydrogen concentration greater than 2% by I I volume, immediately suspend all additions of waste gases to the affected waste gas holdup tank and reduce the concentration of oxygen j to less than or equal to 2% by volume within one hour. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. I c. t t SURVEILLANCE REQUIREMENTS I i 4.11.2.5 The concentration of hydrogen and oxygen in the waste gas holdup l .i -system shall be determined to be within the above limits by monitoring the i waste gas additions to the waste gas holdup system with the hydrogen and v j. Oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10. 4 i i 5 + i . i, i t t 4 4 SEQUOYAH - UNIT'1 3/4 11-15 Amendment No. 12 ~_

E/OIC":TI'JE EFFLUENTS GAS DECAY-TANXS ( v LIMITING CONDITION FOR OPERATION -a 3.11.2.6 The quantity of radioactivity contained in each gas decay tank shall I be limited to less than or equal to 50,000 curies of noble gases (considered as Xe-133). APPLICABILITY: At all times. ACTION: With the quantity of radioactive material in any gas decay tank a. exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit. h. The prcvis ici.3 of 5peci f ications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas decay tank shall be determined to be within the above limit at least once per 24 hours when radioactive materials are being added to the tank. I -= SEQUOYAH - UNIT 1 3/4 11-16

r POWER DISTRIBUTION LIMITS BASES When an F measurement is taken, an allowance for both experimental error q and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance. RTP The F limit for RATED THERMAL POWER (Fxy ) as provided in the Radial xy Peaking Factor limit report per Specification 6.9.1.14 was determined from expected power control maneuvers over the full range of burnup conditions in the core. When RCS flow rate and F are measured, no additional allowances are g necessary prior to comparison with the limits of Figures 3.2-3 and 3.2-4. Measurement errors of 3.5% for RCS total flow rate and 4% for F have been allowed for in determination of the design DNBR value. The 12 hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of op, ration shown on Figure 3.2-3. 3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distri-bution satisfies the design values used in -the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation. The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x y plane power tilts. The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and cor-rection of a dropped or misaligned rod. In the event such action does not 1 correct the tilt, the margin for uncertainty on F is reinstated by reducing q the power by 3 percent for each percent of tilt in excess _ of 1.0. i -SEQUOYAH~- UNIT 1 8 3/4 2-5 Amendment No. 12 m

POWER DISTRIBUTION' LIMITS 3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the para-meters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. -The 12 hour periodic surveillance of these parameters thru instrument c readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. SEQUOYAH - UNIT 1 B 3/4 2-6 Amendment No. 12

t i I 3/4.4 REACTOR COOLANT SYSTEM l BASES i 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant-loops in opera-tion, and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES I and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 1 hour. 4 In MODE 3, a single reactor coolant loop provides sufficient heat removal capabiliity for removing decay heat; however, single failure considerations require that two loops be OPERABLE. In MODE 4, a single reactor coolant loop or residual heat removal (RHR) loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if.the reactor coolant loops are not OPERABLE, this specification i requires two RHR loops to be OPERABLE. In MODE 5, single failure considerations require that two RHR loops be 4 OPERABLE. j The operation of one Reactor Coelant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. 3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve set point. The relief capacity.of a single safety valve is adequate to relieve any over-pressure condition which could occur during shutdown. In the event that no SEQUOYAH - UNIT 1 B 3/4 4-1 Amendment No. 12

4 i REACTOR COOLANT SYSTEM BASES 4 safety valves are OPERABLE, an operating RHR loop, connected to the RCS, i provides overpressure relief capability and will prevent RCS overpressurization. i to t During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. t The combined relief capacity of all of these valves is greater than the maximum -l surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves. l Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. The power operated relief valves (PORVs) and steam bubble function to j relieve RCS pressure during all design transients up to and including the l design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. i Each PORV has a remotely operated block valve to provide positive shutoff capability should a relief valve become inoperable. r 3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the I parameter is maintained within the normal steady state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. i The 12 hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is-formed and thus the i RCS is not a hydraulically solid system. The requirement that 150 kw of pressurizer heaters and their associated controls be capable of being supplied l electrical power from an emergency bus provides assurance that the plant will be able to control reactor coolant pressure and establish natural circulation I conditions. r t 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the kCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the-tubes in the event that there is evidence of mechanical damage or-progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. i i t r I h SEQUOYAH - UNIT 1 B 3/4 4-2 Amendment No. 12 i ,,e ,m,

PLANT SYSTEMS BASES -3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the ~ resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses. 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity ~ effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses. 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70 F and 200 psig are based on a steam generator RT f 60 F and are sufficient to prevent brittle fracture NDT 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this ~ system, assuming a single failure, is consistent with the assumptions used in the accident analyses. 3/4.7.4 ESSENTIAL RAW COOLING WATER SYSTEM The OPERABILITY of the essential raw cooling water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits. t SEQUOYAH - UNIT 1 8 3/4 7-3 Amendment No. 12

t PLANT SYSTEMS BASES 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM The OPERABILITY of the auxiliary building gas treatment system ensures that radioactive materials leaking from the ECCS equipment following a LOCA are filtered prior to reaching the environment. The operation of this system and the resultant effect on offsite dosage calculations was assumed in the accident analyses. ANSI N510-1975 will be used as a procedural guide for ~ surveillance testing. Cumulative operation of the system with the heaters on for 10 hours over a 31 day period is suf ficient to reduce the buildup of moisture on the adsorbers and HEPA filters. 3/4.7.9 SNUBBERS Snubbers are designed to prevent unrestrained pipe or component motion under dynamic loads as might occur during an earthquake or severe transient, while allowing normal thermal motion during startup and shutdown. The con-sequence of an inoperable snubber is an increase in the probability of structural damage to piping or components as a result of a seismic or other event initiating dynamic loads. It is therefore required that all snubbers required.to protect the primary coolant system or any other safety system or component be operable during reactor operation. Because the snubber protection is required only during relatively low probability events, a period of 72 hours is allowed to replace or restore the-inoperable snubber (s) to operable status and perform an engineering evaluation on the supported component or declare the supported system inoperable and follow the appropriate limiting condition for operation statement for that system. The engineering evaluation is performed to determine whether the mode of failure of the snubber has adversely affected any safety-related component or system. Safety-related snubbers are visually inspected for overall integrity and operability. The inspection will include verification of proper orientation, adequate fluid level if applicable, and attachment of the snubber to its anchorage. TLe removal of insulation or the verification of torque values for ~ threaded fasteners is not required for visual inspections. The inspection frequency is based upon maintaining a constant level of snubber protection. Thus, the required inspection interval varies inversely with the observed snubber failures. The number of inoperable snubbers found during a required inspection determines the time interval for the next required inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25 percent) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule. When the cause of the rejection of a snubber in a visual inspection is clearly established and remedial for that snubber and for any other snubbers SEQUOYAH - UNIT 1 B 3/4 7-5 Amendment No.12

PLANT SYSTEMS BASES SNUBBERS (Continued) that may be generically susceptible and operability verified by inservice func-tional testing, if applicable, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have the same design features directly related to rejection of the snubber, or are similarly located or exposed to the same environmental conditions such as temperature, radiation, and vibration. Inspection groups may be ?stablished based on design features and installed conditions which may be exsected to be generic. Each of these inspection groups are inspected and tested separately unless an engineering analysis indicates the inspection group is improperly constituted. All suspect snub-bers are subject to inspection and testing regardless of inspection groupings. To further increase the assurance of snubber reliability, functional tests shall be performed during each refueling outage. These tests will include stroking of the snubbers to verify proper movement, activation, and bleed or release. The performance of hydraulic snubbers generally depends on a clean, deaerated fluid contained within variable pressure chambers, flowing at closely controlled rates. Since these characteristics are subject to change with expo-sure to the reactor environment, time, and other factors, their performance within the specified range should be verified. Mechanical snubbers which depend upon overcoming the inertia of a mass and the braking action of a capstan spring contained within the snubber for limiting the acceleration of the attached compo-nent (within the load rating of the snubber) are not subject to changes in per-formance in the same manner as hydraulic snubbers. Pending the development of information regarding the change during the service of the snubber of the acceleration / resistance relationship and the optimum method for detecting this change, these mechanical snubbers may be tested to verify that when subjected to a large change in velocity the resistance to movement increases greatly. The performance change information is to be developed in order to establish test methods to be used during and after the first refueling outage. Ten percent of the total population of approximately 700 snubbers is an adequate sample for functional tests. The initial sample is to be proportioned. among the groups in order to obtain a representative sample. Observed failures of more than two snubbers in the initial lot will require an engineering analysis and testing of additional snubbers selected from snubbers likely to have the same defect. A thorough inspection of the snubber threaded attach-ments to the pipe'or components and the anchorage will be made in conjunction with all required functional tests. 3/4.7.10 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak test-ing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with surveil-lance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continu-ously enclosed within a shielded mechanism (i.e., sealed sources within radia-tion monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism. i SEQUOYAH - UNIT 1 B 3/4 7-6 Amendment No. 12 )

3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1 and 3/4.8.2 A.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the A.C. and D.C power sources and associated distri-bution systems during operation ensures that sufficient power will be available to supply the safety related equipment required for 1) the safe shutdown of the facility and 2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criteria 17 of Appendix "A" to 10 CFR 50. The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commen-surate with the level of degradation. The OPERABILITY of the power sources are consistent with the initial condition assumptions of the accident analyses and are based upon maintaining at least one redundant set of onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of the other onsite A.C. source. The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that 1) the facility can be maintained in the shutdown or refueling condition for extended time periods and 2) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status. The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guides 1.9 " Selection of Diesel Generator Set Capacity for Standby Power i' Supplies," March 10, 1971, and 1.108 " Periodic Testing of Diesel Generator i Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977, and 1.137 " Fuel-Dil Systems for Standby Diesel Generators," Revision 1, October 1975. The Surveillance Requirement for demonstrating the OPERABILITY of the Station batteries are based on the recommendations of Regulatory Guide 1.129 " Maintenance Testing and Replacement of Large Lead Storage Batteries for ] Nuclear Power Plants," February 1978, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large lead Storage Batteries for Generating Stations and Substations." Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage onfloat charge, connection resistance values and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity. Table 4.8-2 specifies the normal limits for each designated pilot cell 8 and each connected cell for electrolyte level, float voltage and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal SEQUOYAH - UNIT 1 B 3/4 8-1 Amendment No. 12

ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued) limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more thar. 010 below the manufacture's full charge specific gravity, ensures the OPERABILITY and capability of the battery. 4 Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2 is permitted for up to 7 days. During this 7 day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than.020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than.040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function. 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors are protected by either de energizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protection circuit b.eakers during periodic surveillance. The surveillance requirements applicable to lower voltage circuit breakers and fuses provides assurance of breaker and fuse reliability by testing at least one representative sample of each manufacturers brand of circuit breakers and/or fuse. Each manufacturer's molded case and metal case circuit breakers and/or fuses are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers and/or fuses are tested. If a wide variety exists within any manufacturer's brand of circuit breakers and/or fuses, it is necessary to divide that manufacturer's breakers and/or fuses into groups and treat each group as a separate type of breaker of fuses for surveillance purposes. The OPERABILITY of the motor operated valves thermal overload protection and/cr bypass devices ensures that these devices will not prevent safety related valves from performing their function. The Surveillance Requirements for demonstrating the OPERABILITY of these devices are in accordance with Regulatory Guide 1.106 " Thermal Overload Protection for Electric Motors on Motor Operated Valves", Revision 1, March 1977. Circuit breakers actuated by fault currents are used as isolation devices in this plant. The OPERABILITY of these circuit breakers ensures that the 1E busses will be protected in the event of faults in nonqualified ioads powered by the busses, f a SEQUOYAH - UNIT 1 B 3/4 8-2 Amendment No. 12

RADI0 ACTIVE EFFLUENTS BASES ? 3/4.11.2.6 GAS DECAY TANKS Restricting the quantity of radioactivity contained in each gas decay -tank provides assurance that in the event of.an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 15.7.1, " Waste Gas System Failure". 3/4.11.3 SOLID RADI0 ACTIVE WASTE The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in estuolishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times. 3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result in the limitation of dose to a member of the public for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. I 1 f l l l l I i SEQUOYAH - UNIT I B 3/4 11-5 Amendment No. 12 L

DESIGN FEATURES i 5.6 FUEL STORAGE CRITICALITY - SPENT FUEL

5. 6.1.1-The spent fuel storage racks are designed and shall be maintained with:

a. Ak equivalent to less than 0.95 when flooded with unborated eff water, which includes a conservative allowance of 1.78% delta k/k for uncertainties as described in Section 4.3 of the FSAR. b. A nominal 10.375 inch center-to-center distance between fuel assemblies placed in the storage ranks. CRITICALITY - NEW FUEL 5.6.1.2 The new fuel pit storage racks are designed and shall be maintained with a nominal 21.0 center-to-center distance between new fuel assemblies such that k will n t exceed 0.98 when fuel having a maximum enrichment of 3.5 eff weight percent U-235 is in place and aqueous foam mcderation is assumed. DRAINAGE 5.6.2 The spent fuel pit is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 722 ft. CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a i storage capacity limited to no more than 1386 fuel assemblies. 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1. i. l SEQUOYAH - UNIT 1 5-5 Amendment No. 12

TABLE 5.7.1 h COMPONENT CYCLIC OR TRANSIENT LIMITS = 0

d. ~

CYCLIC OR DESIGN CYCLE COMPONENT TRANSIENT LIMIT OR TRANSIENT 1

1 Reactor Coolant System 200 heatup cycles at < 100 F/hr Heatup cycle - T from < 200 F j'

and 200 cooldown cycles at to > 550 F.- avg i < 100 F/hr Cooldown cycle - T from > 550 F to < 200 F"V9 a 200 pressurizer cooldown cycles Pressurizer cooldown cycle at < 200*F/hr 4 temperatures from > 650*F to < 200*F. 80 loss of load cycles, without > 15% of RATED THERMAL POWER to immediate turbine or reactor trip. 0% of RATED THERMAL POWER. 40 cycles of-loss of offsite Loss of offsite A.C. electrical A.C. electrical power. en a power source supplying the onsite ESF Electrical System. 80 cycles of loss of flow in one loss of only one reactor reactor coolant loop. coolant pump. 400 reactor trip cycles. 100% to 0% of RATED THERMAL POWER. 10 inadvertent auxiliary spray Spray water temperature differential actuation cycles. > 320"F. 50 leak tests Pressurized to 2485 psig 5 hydrostatic pressure tests Pressurized to 3105 psig Secondary System 5 hydrostatic pressure tests Pressurized to 1330 psig 'N [ ] g"'N ) O n

TABLE 6.1-1 SURVEILLAtCE REQUlREMENTS PERE 0RMED BY RADIOLOGICAL HYGIENE BRANCH d 4.11.1.2 4.11.1.3.1 4.11.2.1.1 (partial) 4.11.2.1.2 (partial) 4.11.2.2 4.11.2.3 4.11.2.4 4.11.4 4.12.1 4.12.2 4.12.3 a SEQUOYAH - UNIT 1 6-5 N

~ Table 6.2-1 Minimum shift crew composition With Unit 2 in Mode 5 or 6 or De-fueled Position . Number of individuals required to fill position Modes 1, 2, 3, & 4 Modes 5 & 6 a a SS l y SRO 1 None R0 2 1 b A0 2 2 STA 1 None With Unit 2 in Modes 1, 2, 3, or 4 Po s',i on Number of individuals required to fill position Modes 1, 2, 3, & 4 Modes 5 & 6 a a SS l y a SR0 l None D R0 2 1 b A0 2 y t STA 1" None aIndividual may fill the same position on Unit 2 b0ne of the two required individuals may fill the same position on Unit 2 SS - Shift Supervisor with a Senior Reactor Operators License on Unit 1 SR0 - Individual with a Senior Reactor Operators License on Unit 1 R0 - Individual with a Reactor Operators License on Unit 1 A0 - Auxiliary Operator STA - Shift Technical Advisor Except for the Shift Supervisor, the Shift Crew Composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shif t crew members provided immediate action is taken to restore the Shif t Crew Composition to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent. During any absence of the Shift Supervisor from the Control Room while the unit is in MODE 1, 2, 3 or 4, an individual (other than the Shift Technical Advisor) with a valid SR0 license shall be designated to assume the Control Room command function. During any absence of the Shift Supervisor from the Control Room while the Unit is in Mode 5 or 6, an' individual with a valid SRO or R0 license (other than the Shift Technical Advisor) shall be designated to assume the Control Room command function. SEQUOYAH -' UNIT 1 6-6 Amendment No.12 m-

ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG) FUNCTION 6.2.3.1 The ISEG shall function to examine plant operating characteristics, ~ NRC issuances, industry advisories, Licensing Event Reports and other sources which may indicate areas for improving plant safety. COMPOSITION 6.2.3.2 The ISEG shall be composed of at least five dedicated full-time engineers located onsite. RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of plant activities to provide independent verification

that these activities are performed correctly and that human errors are reduced as much as practical.

AUTHORITY 6.2.3.4 The ISEG shall make detailed recommendations for revised procedures, equipment modifications, or other means of improving plant safety to the Assistant Director for Maintenance and Engineering of the Division of Nuclear Power. 6.2.4 SHIFT TECHNICAL ADVISOR (STA) 6.2.4.1 The STA shall serve in an advisory capacity to the shif t supervisor on matters pertaining to the engineering aspects of assuring safe operation of the unit.

6. 3 UNIT STAFF QllALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifi-cations of ANSI N18.1-1971 for comparable positions and the supplemental requirements specified in Section 4 and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, except for the Health Physicist who shall meet or

~ exceed the qualifications of Regulatory Guide 1.8, September 1975.

Not responsible for sign-off function.

SEQUOYAH - UNIT 1 6-7 Amendment No. 12

1 ADMINISTRATIVE CONTROLS r (iv) Procedures for the recording and management of data, l (v) Procedures defining corrective actions for off control point i chemistry ' conditions, (vi) Procedures identifying (a) the authority responsible for the interpretation of the data; and (b) the sequence and timing of administrative events required to initiate corrective action, and (vii) Monitoring of the condensate at the discharge of the condensate pumps for evidence of condenser in-leakage. When condenser in-leakage is confirmed, the leak shall be repaired, plugged, or isolated within 96 hours. I d. Backup Method for Determining Subcooling Margin i A program which will ensure the capability to accurately monitor the ] Reactor Coolant System Subcooling Margin. This program shall include j the following: i (i) Training of personnel, and (ii) Procedures for monitoring. e. Postaccident Sampling A program which will ensure the capability to obtain and analyze j reactor coolant, radioactive iodines _and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following: i (i) Training of personnel, (ii) Procedures for sampling and analysis, (iii) Provisions for maintenance of sampling and analysis equipment. 4 4 l t i i SEQUOYAH - UNIT 1 6-15b Amendment No.12 I

I ADMINISTRATIVE CONTROLS (.- 6.9 REPORTING REQUIREMENTS ' ~ ROUTINE REPORTS AND REPORTABLE OCCURRENCES d

6. 9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Of fice of Inspection and Enforcement unless otherwise noted.

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power levei, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic perf ormance of the plant.

6. 9.1. 2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report. 6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed. s SEQUOYAll - UNIT 1 6-16 l (

t.DMINISTRATIVE CONTROLS L.. ~ v h. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specifications that have or could have permitted reactor operation in a manner less conservative j than assumed in the analyses. i. Performance of structures, systems, or components that requires l remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical specifications bases; or 4 discovery during unit life of conditions not specifically considered in the safety analysis report or technical specifications that require remedial action or corrective measures to prevent the j existence or development of an unsafe condition. 'j. Offsite releases of radioactive materials in liquid and gaseous ef fluents which exceed the limits of Specification 3.11.1.1 or 3.11.2.1. k. Exceeding the limits in Specification 3.11.1.4 or 3.31.2.6 for the i storage of radioactive materials in the listed tanks. The written follow-up report shall include a schedule and a description of l activities planned and/or taken to reduce the contents to within the 'specified limits. [ THIRTY DAY WRITTEN REPORTS 6.9.1.13 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty days of occur-rence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event. Reactor protection system or engineered safety feature instrument a. i settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfill-ment of the functional requirements of affected systems. b. Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation. c. Observed inadequacies in the implementation of administrative or i procedural controls which threaten to cause reduction of degree of l redundancy provided in reactor protection systems or engineered safety feature systems. d. Abnormal degradation of systems other than those specified in 6.9.1.12.c above designed to contain radioactive material resulting from the fission process. SEQUOYAH - UNIT 1 6-21 ~

ADMINISTRATIVE CONTROLS e. An unplanned offsite release of 1) more than I curie of radioactive 4 material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following information: f 1. A description of the event and equipment involved. 2. Cause(s) for the unplanned release. 3. Actions taken to prevent recurrence. 4. Consequences of the unplanned release. f. Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 3.12-2 when averaged over any calendar quarter sampling period. I RADIAL PEAKING FACTOR LIMIT REPORT limit for Rated Thermal Power (F,RTP) shall be provided to 6.9.1.14 The F the Director of the Regional Office of Inspection and Enforcement, with a copy to the Director, Nuclear Reactor Regulation, Attention, Chief of the Core Performance Branch, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 for all core planes containing bank "D" control rods and all unrodded core planes at least 60 days prior to cycle initial criticality. In the event that the limit would be submitted at some other time during core life, it will be i submitted 60 days prior to the date the limit would become effective unless i otherwise exempted by the Commission. RTP j Any information needed to suport F will be by request from the NRC and need not be included in this report. V SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. 6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. 6.10.1 The following records shall be retained for at least five years: Records and logs of unit operation covering time interval at each a. power level. b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety. All REPORTABLE OCCURRENCES submitted to the Commission. c. d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications. Records of changes made to the procedures required by Specification c. 6.8.1 and 6.8.4. f. Records-of radioactive shipments, g. Records of sealed source and fission detector leak tests and results. h. Records of annual-physical inventory of all sealed source _ material of record. SEQUOYAli UNIT 1 6-22 Amendment No.12

s ADMINISTRATIVE CONTROLS 4 6.10.2 The following records shall be retained for the duration of the Unit Operating License: Records and drawing changes reflecting unit design modifications a. made to systems and equipment described in the Final Safety Analysis Report. b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories. j c. Records of radiation exposure for all individuals entering radiation i control areas. d. Records of gaseous and liquid radioactive material released to the f

environs, Records of transient or operational cycles for those unit components e.

identified in Table 5.7-1. i f. Records of reactor tests and experiments. 4 g. Records of training and qualification for current members of the unit staff. h. Records of in-service inspections performed pursuant to these j Technical Specifications, i. Records of Quality Assurance activities required by the Operational Quality Assurance Manual. j. Records of reviews performed for changes made to procedures or i equipment or reviews of tests and experiments pursuant to 10 CFR 50.59. k. Records of meetings of the PORC, RARC, and the NSRB. 1. Records of analyses required by the radiological environmental j monitoring program. Records of secondary water sampling and water quality. m. Records of the service live monitoring of all hydraulic and mechanical n. i snubbers listed on Tables 3.7-4a and 3.7-4b, including the maintenance i performed to renew the service life. Records for Environmental Qualification which are covered under the o. ~ j provisions of Paragraph 2.c.(12)(6) of License No. DPR-77. 6.11 RADIATION PROTECTION PROGRAM ' Procedures for personnel radiation protection shall be prepared consistent avith the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. h 1 f SEQUOYAH - UNIT 1 6-23 Amendment No. 12

ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c) (2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Special (Radiation) Werk Permit

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

A radiation monitoring device which continuously indicates the a. radiation dose rate in the area. b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made af ter the dose rate level in the area has been established and personnel have been made knowledgeable of them. An individual qualified in radiation protection procedures wno is c. equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Special (Radiation) Work Permit. ( 6.12.2 The requirements of 6.12.1, above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shif t Engineer on duty and/or the Health Physicist. 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 - The PCP shall be approved by the Commission prior to implementation. 6.13.'> Licensee initiated changes to the PCP: 1. Shall be submitted to the Commission in the semi annual Radioactive Etfluent Release Report for the period in which the change (s) was This submittal shall contain: made. suf ficiently detailed information to totally support the rationale a. f or the change without benefit of additional or supplemental information, iHealth Physics personnel or personnel escorted by Health Physics personnel in accordance with approved emergency procedures, shall be exempt from the SWP issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas. {' SEQUOYAH - UNIT 1 6-24 v}}

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c 3 sow

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ML20054D406

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::nn :::~.= =:

==:nr._ t

..j_r.

. + - =;nn-nu t=

c 3 sow

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